safe operation of critical assemblies and research reactors safety standards/safety... ·...

i fSAFETY $ S E R I E S

~ * 7 ' X

No.4

Safe Operation of Critical Assemblies

and Research Reactors

I N T E R N A T I O N A L ATOMIC ENERGY AGENCY V I ENNA 1961

This publication is not longer valid Please see http://www-ns.iaea.org/standards/

SAFE O P E R A T I O N OF C R I T I C A L A S S E M B L I E S A N D R E S E A R C H

R E A C T O R S


The following States are Members of the International Atomic Energy Agency:AFGH AN ISTANALBANIAARG EN TIN AAUSTRALIAAUSTRIABELGIU MBRAZILBU LGARIABURM ABYELORUSSIAN SOVIET

SOCIALIST REPUBLIC CAM BO D IA C A N A D A CEYLON CH ILE CH IN A CO LO M B IA CU BACZECH O SLO VAK

SOCIALIST REPUBLIC DEN M ARKD O M IN IC A N REPUBLICE CU AD O RE L SALVADORETH IO PIAF IN LA N DFRAN CEF E D E R A L REPUBLIC

O F GERM ANY GH AN A G REECE G U A TE M A LA H AITI H O LY SEE H ONDURAS HUNGARY IC E L A N D IN D IA IN D O N ESIA IRAN

ISRAELITALYJAPANREPUBLIC OF KOREALU XEM BO U RGM EXICOM O N ACOM O RO CCON ETH ERLAN D SN E W ZE A LA N DNICARAGU ANORW AYPAKISTANPARAGUAYPERUPHILIPPINESPO LAN DPO RTU G ALRO M AN IASENEGALSPAINSUDANSW ED ENSW ITZE R LA N DTH A ILA N DTUN ISIATURKEYUKRAINIAN SOVIET SOCIALIST

REPU BLIC U N IO N O F SOUTH AFR ICA UN IO N O F SO VIET SOCIALIST

REPUBLICS U N IT E D ARAB REPUBLIC U N IT E D K IN G D O M OF GREAT

BRITAIN A N D NORTH ERN IRELAN D

U N IT E D STATES OF AM ERICA VEN E ZU E LA V IE T-N A M YU GOSLAVIA

IRAQThe Agency’s Statute was approved on 26 O ctober 1956 at an international

conference held at United Nations headquarters, N ew York, and the Agency came into being when the Statute entered into force on 29 July 1957. The first session o f the General Conference was held in Vienna, Austria, the permanent seat o f the Agency, in October, 1957.

The main objective of the Agency is “ to accelerate and enlarge the con tribution o f atomic energy to peace, health and prosperity throughout the world” .

© I A E A , 1961Permission to reproduce or translate the information contained in this publication may be obtained by writing to the International Atomic Energy Agency, Kaerntnerring 11, Vienna I.

Printed in Austria by Paul Gerin, Vienna May 1961


S A F E T Y S E R IE S No. 4

SAFE OPERATION OF CRITICAL ASSEMBLIES AND RESEARCH REACTORS

IN TE R N A TIO N A L ATO M IC ENERGY AGENCY Kaemtnerring, Vienna I, Austria

1961


TH IS M AN U A L IS ALSO PUBLISH ED

IN FREN CH , RUSSIAN A N D SPANISH

SAFE O PERATIO N O F C R IT IC A L ASSEM BLIES AN D RESEARCH REACTORS, IAEA, VIENNA, 1961

STI/PUB/29


F O R E W O R D

The present Manual has been prepared by several o f the *\gency’s reactor specialists in close consultation with an international panel of

the emphasis throughout is on safety, the Manual is published as an issue of the “ Safety Series” .

The contents o f the Manual are in no sense intended to be mandatory; they should be regarded as a guide, based on the best available experience in a number o f countries, to the operation o f critical assemblies and research reactors dn such a way as not to endanger either the operators or the public.

The Manual will no doubt find its widest appeal in the less developed countries where the types o f assembly in question are being operated for the first time; but workers in technologically more advanced parts of the world will also find it valuable as a synthesis o f the best practices in a number of countries.

experts on safe operation of critical assemblies and research reactors. Since

Director General

May 1961


CONTENTS1. I N T R O D U C T I O N .................................................................................... 9

1.1. Purpose .............................................................................................................. 91.2. S c o p e ....................................................................................................................... 91.3. Limitations ..................................................................................................... 121.4. D e fin it ion s .............................................................................................................. 13

2. C R I T I C A L A S S E M B L I E S ....................................................................152.1. Safety o f Design o f Critical A s s e m b lie s ................................................ 152.2. Safety in Instrumentation o f Critical Assemblies ..............................212.3. Safety o f Construction o f Critical A s s e m b lie s .......................................272.4. Safety in Operation o f Critical A ssem blies ............................................... 272.5. Radiological Protection Instrumentation ................................................35

3. R E S E A R C H R E A C T O R S ....................................................................383.1. Safety of Design and Instrumentation o f Research Reactors . . . . 383.2. Commissioning of Research R e a c t o r s ........................................................ 413.3. Safety in Operation of Research R e a c t o r s ................................................423.4. Experimental Use o f Research Reactors ................................................473.5. Radiological Protection Instrumentation ................................................50

4. P E R S O N N E L .......................... ............................................................ 514.1. Introduction ..................................................................................................... 514.2. Qualifications ..................................................................................................... 514.3. On-Site Training ............................................................................................ 524.4. Examination and Control ................................................ .....................53

5. A D M I N I S T R A T I V E P R O C E D U R E S ........................................... 545.1. The Safety Committee System ................................................................. 545.2. Duties o f the Safety C o m m it t e e ................................................................. 565.3. Composition o f the Safety Committee ................................................565.4. Safety Committee P ro ce d u re .......................................................................... 585.5. Duties o f the Operations Safety Committee ....................................... 595.6. Composition of the Operations Safety C o m m itte e .............................. 605.7. Operations Safety Committee P r o c e d u r e ................................................61

6. S A F E T Y D O C U M E N T S ....................................................................636.1. Safety R e p o r t s ..................................................................................................... 636.2. Form for Routine Irradiations .................................................................. 686.3. Form for Simple Experiments .............................. ............................. 706.4. Core Certificate ............................................................................................ 706.5. Other Documentation ................................................................................... 73


7. E M E R G E N C Y P R O C E D U R E S .................................................................747.1. General .................................................................................................... 747.2. Local I n c id e n t ................................................................................................... 757.3. Site E m erg en cy ...................................................................................................777.4. Public E m e r g e n c y .......................................................................................... 797.5. Requirements of the Health Physics S e r v i c e ....................................... 80

A C K N O W L E D G E M E N T ............................................................................ 81

A P P E N D I X I: B I B L I O G R A P H Y .................................................................82Section A: General References ........................................................................82Section B: Safety Reports .................................................................................85Section C: Siting and Containment ............................................................... 85

A P P E N D I X II: R E A C T O R S T A N D I N G O R D E R S A N DO P E R A T I N G I N S T R U C T I O N S ........................................................87Section A: Reactor Standing O r d e r s ...............................................................87Section B: Reactor Operating Instructions...................................................... 92

A P P E N D I X III: S Y L L A B U S OF A R E A C T O R T R A I N I N G C O U R S E ......................................................................................................................95


1. INTRODUCTION

1 . 1 .

1.1.1.

1.1.2 .

1.1.3.

1.2 .

1.2 .1.

PURPOSE

This Manual is provided as a guide to the safe operation, as defined in Section 1.4.4., o f critical assemblies and small research reactors. It is intended that it should be used by all authorities and persons concerned with, or responsible for, the use of such equipment, in addition to the scientists and technologists who are actually working with, or operating it. It is suggested that it will be o f use to those wishing to design and manufacture, or purchase, critical assemblies or research reactors, as well as those already in possession of them, and that it will prove particularly helpful to those users who have no direct access to other collected sources of information.This Manual is not a set of rules or a code of practice, but a series of recommendations which must be interpreted with scientific judgement in their application to any particular problem. The guiding principles are given from which good operational procedures may be established and improved. The promulgation o f rigid standards is both impossible and undesirable at the present time, since the topics discussed form part o f a rapidly growing science and technology. Therefore, any recommendations made should not be used to restrict or inhibit future developments.The Manual is intended mainly for use in those M ember States where there has been little experience in the operation of critical assemblies and research reactors. It has been com pounded from the best practices which exist in M ember States having a large amount of such experience, so that nothing in it should conflict with the best practices to be encountered in the field of safe operation.

SCOPE

The safe operation o f critical assemblies and research reactors involves very many fields of information and experience. M any o f these are covered in this Manual. Not all o f them are directly concerned with safe operation. The fields included in this Manual are the administrative procedures required, safety in design and construction, the commissioning o f critical assemblies and research reactors, staff training, qualifications and experience, the operation of such equipment, and (especially for research reactors) their

!>


experimental use. A n Appendix* describes the detailed application o f the Manual to a hypothetical research reactor, and a Bibliography** is also included.

1.2.2. The ultimate safety of critical assemblies and research reactors can only be ensured if due attention is given to safety at all stages in' the life o f the critical assembly or reactor. The design and material selection must meet acceptable standards and must incorporate such special safety features as are judged necessary; the construction must be of the requisite quality and use the material specified; the operation of the system must take into account the possibility of accident due to error or to the system itself; and finally, experiments perform ed with the aid o f the system must be designed and carried out so that neither the isystem nor the experiment is in any way endangered.

1.2.3. For the purpose of this Manual, some o f the subject-matter hasbeen divided into two parts. This division occurs naturally because some systems require separate operating and experimental teams for their efficient use, whilst others may be better utilized by means of a single team. However, in the latter case, particular personnel within the team may be primarily responsible for operation or experimental use.

1.2.4. Section 2, headed “ Critical Assemblies” , includes discussion ofthe safety of critical assemblies themselves, zero-energy reactor systems (i. e. those operating at pow er levels up to a few watts), and so-called sub-critical assemblies when sufficient fuel and moderator is present to enable criticality to be achieved, even if only accidentally. All these systems are usually utilized by a single team.

1.2.5. Section 3 o f the Manual, entitled “ Research Reactors” , will beapplicable to small- and medium-sized research reactors primarily used by physicists rather than by engineers involved in large- scale experiments. It will be applicable also to reactors which have facilities for small-scale engineering experiments, but it is not intended that this Section, or the Manual, w ould cover those high-power, high-flux reactors which are used for large-scale engineering experiments. It is difficult to set an upper size limit in terms o f pow er level, since this is very dependent upon the reactor type. It is anticipated that the scope of this Section will

Appendix II, p. 87. Appendix I, p. 82.


include those reactors operating in the kilowatt region, and may include some reactors operating in the megawatt range. It is expected that these reactors will be staffed by separate teams for operation and experimental use.

1.2.6. In the Section headed “ Critical Assemblies ” , design and construction have been discussed in some detail, but for the “ Research Reactor” Section the general principles only have been given, although attention has been drawn to the design o f experimental equipment. The design and construction o f a research reactor is necessarily a com plex procedure requiring a large and highly- experienced design organization, together with w ell-developed manufacturing facilities. The details o f design and construction o f research reactors are felt, therefore, to be beyond the scope of this Manual.

1.2.7. A Section of the Manual is devoted to emergency procedures. This Section is not to be regarded as a com plete guide to such procedures. H owever, those sections o f this work which are necessary to minimize hazards in the event o f an untoward incident with the assembly or reactor are considered. Similarly, the Manual is not a com plete guide to the radiological protection organization of an establishment, although this subject has been discussed where it is necessary.

1.2.8. The Section concerning training is intended to serve as a guide to the qualifications and experience which are to be expected of experimental and operating staff at the various levels. The training methods are also briefly discussed.

1.2.9. One Section has been devoted to administrative procedures. This is regarded as a most important Section, since good organization can prevent many accidents. Experience to date has shown that most reactor accidents, and especially those occuning in critical assemblies, might have been avoided if adequate written proce- duces had been available and adhered to, and if sufficient emphasis during training had been placed upon the requirements o f personal discipline during operations. This Manual is primarily concerned with nuclear safety and not with what might be termed “ conventional safety” . It is assumed, therefore, that administrative procedures governing conventional safety will be carried out as has been the practice in the past, and in accordance with national legislation. Examples o f such matters are safe design o f pressure vessels; inspection o f such vessels and piping during and after manufacture; mandatory building require-

11


1.2.10.

1.3.

1.3.1.

1.3.2.

1.3.3.

1.3.4.

12

ments; and regulation of operations with explosive or flammable materials.Three appendices are included. The first contains a bibliography of freely available reports, and is divided into three sections. Section A contains general references and references to operating and administrative procedures. Section B contains references to safety reports, whilst Section C is a bibliography of important references to the problems o f siting and containment (Paragraph 1.3.1.). The second appendix provides examples of reactor standing orders and operating instructions, involving the application of the principles contained in this manual to a particular reactor system. A third appendix gives a suggested syllabus for a reactor training course.

LIM ITATIO N S

The subject of siting research reactors has not been considered in this Manual. Similarly, the related problem of containment o f reactors has not been discussed. These two omissions are deliberate and have been made for two reasons. Firstly, the basic approach to these subjects is not yet clearly defined.' There are considerable differences in practice between M em ber States. M uch work remains to be done in this field before an international series o f recommendations could be produced. Secondly, these two questions are at present a matter for national policy and are associated with the problems of national licensing procedures and control o f reactor siting. These problems should form the subject o f a separate manual. It has been assumed in this Manual that the site for the reactor has been chosen and that the question of containment has been agreed. Because of the importance o f these problems, a bibliography of freely available documents has been included as Section C of Appendix I. Research reactors used in fast kinetic experiments (such as BORAX) have not been included in this Manual. Such reactors from a very small class o f their ow n; their design and operation are intimately connected with their siting and containment. Often they are designed in the knowledge that the experiment to be perform ed on them in order to obtain certain dynamic information might lead to their self-destruction.Fast-neutron critical assemblies are excluded from discussion in this Manual.Similarly, critical assemblies primarily used for the determination


1.3.6.

1.4.

1.4.1.

.1.4.2.

1.4.3.

1.4.4.

1.4.5.

o f physical parameters for use in chemical plant design have not been included in this Manual. H owever, it is hoped that the guiding principles laid down may prove useful in this connection.The question of radiation exposure of personnel has not been dealt with in this Manual. For details o f maximum permissible occupational exposures, reference should be made to the Recommendations of the International Commission for Radiological Protection (ICRP) published in 1954, and subsequently amended in 1956 and 1958. It is strongly recom m ended that these recommendations be adopted. Some portions of them are summarized in “ Safe Handling of Radioisotopes” , IAEA, Vienna, 1958 (STI/PU B/1).Conventional safety requirements have not been dwelt upon in any great detail since in most cases these are already the subject of national standards or regulations, and may becom e subject to discussion by international organizations such as the International Organization for Standardization (ISO).

D EFIN IT IO N S

(The definitions which follow are included in order to clarify the contents o f this Manual, and are not necessarily exhaustive.)A subcritical assembly is an assembly o f nuclear fuel and m oderator where insufficient fuel and moderator is present to enable criticality to be achieved, even if only accidentally.A critical assembly is a neutron multiplying system which is flexible in character, assembled from fissile and other materials used in nuclear techniques, and used mainly for the determination of critical mass and other characteristics o f the assembly or materials.A research reactor is a more permanent system used mainly for the generation of neutron flux and ionizing radiations used for research purposes and irradiation of materials.Safe operation is defined as operation in accordance with recognized nuclear and conventional safety procedures in order to provent hazard to any person.Reactivity, A k/k, is defined as the value of the expression (keft — l)//ccfF, where Avtf is the effective multiplication constant. Reactivity is a measure o f the departure o f a reactor or a portion o f it from the critical condition.

IB


1.4.6.

1.4.7.

1.4.8.

A local incident is one confined to one building or part of it and means any deviation from the foreseen operational conditions involving radioactive materials which may imply a hazard for personnel or equipment.A site emergency means any incident involving escape of radioactive materials from the local area or the presence of ionizing radiations which may constitute a hazard to personnel or property within the boundaries of the establishment.A public emergency is any set of circumstances involving radioactive materials which may cause hazards or damage to persons or property outside the boundaries of the establishment.

14


2. CR ITIC AL ASSEMBLIE S

This Section of the Manual, which is intended only for the reader who has had previous experience in the nuclear field (but not necessarily of critical assemblies), is devoted to the specification, design and construction of critical assemblies as well as their actual operation. The advice of the appropriate Safety Committee should be sought as described in Section 5, whilst safety reports should be written as described in Section 6. The operation of such an assembly is intimately connected with its specification and design, and it is not possible completely to separate the two. Fast-neutron critical assemblies are not covered by this Section, since it is not expected that they would be designed and operated by users of this Manual. The essential dangers inherent in critical assemblies but not found in research reactors arise from the element of uncertainty of the critical parameters, the large excess reactivities which may be possible and the frequent direct access to fuel elements. The designer and manager of a critical experiment have always to be aware that the expected calculated values may be in error, and may differ greatly from the actual values. The safety of the system has to be assured much more by instrumentation and safety devices than by calculated values. Neither safety circuits, interlocks, mechanical design nor administrative control can possibly be a sufficient condition to ensure the safe operation of a critical assembly or a research reactor. There can be no substitute for intelligence, alertness, discipline and training of the operating staff.

2.1. SAFETY OF DESIGN OF CRITICAL ASSEMBLIES2.1.1. Reactor physics considerations

The specifications of a critical assembly should, if possible, include estimates of the excess reactivity in the system, values of the maximum rate of reactivity addition, temperature coefficients, void coefficients, speed of heat transfer from fuel to moderator, etc. The physicist in charge must specify these quantities as well as the geometrical layout of the cores and rod removal speeds corresponding to predetermined safe reactivity addition rates. These specifications must be incorporated in the design.

2.1.1.1. For normal static or dynamic critical assemblies, or for one particular lattice in reactors having variable lattices, the maximum excess reactivity must in no case exceed the amount that can

15


be compensated by the permanent available shut-down system, nor must it be more than that estimated to be required for the experiment, and the design and construction must be such as to ensure that accidental increases of reactivity due to mechanical failures are impossible.

2.1.1.2. Reactivity additions to a system when it is not in the shut-down condition should be in discrete amounts, each addition being followed by a period of time during which the effect of the addition may be observed. The design should incorporate features to ensure this. The maximum value of such an addition should be less than that required to cause prompt critical conditions when added to the assembly at the point of criticality, since under these conditions the effectiveness of the control devices may be inadequate. The assembly must be designed so that each addition of reactivity requires further deliberate operational action before a further separate addition can be made.

2.1.1.3. The rate of reactivity addition during the critical approach procedure must be limited so that the safety features of the system have time to become fully effective before unsafe conditions are reached. The value normally used is 2 X 1 0 -20/o A k/k per sec at any time. When fuel additions are made by hand during the critical approach experiment with one bank of shut-down rods within the reactor, the subsequent withdrawal of negative reactivity should also be limited to the same value.

2.1.1.4. When the additions of reactivity are made by increasing some fundamental parameter which cannot be varied in sufficiently fine increments or in a sufficiently accurately controlled way (e. g. fuel loading, moderator height), a variable control parameter (e. g. moderator height, control rod) is also necessary. In these cases, the control parameter must be designed to prevent criticality ever being achieved during the increase in fundamental parameter; the incremental additions of fundamental parameter must be limited to less than the worth of the control parameter; and the limiting conditions of subparagraphs 2.1.1.2. and 2.1.1.3. will apply to the control parameter.

2.1.1.5. The design should incorporate features to permit the approach to criticality to be achieved within a reasonable period of time. To this end the values of both the addition and the rate of addition may be increased during the first part of the addition operation. These increased values must not be used beyond a safe limit. The value of this limit is difficult to define generally,

1 6


as it will depend upon the nature of the assembly, and the parameter which is variable (e. g. mass, moderator height, etc.). It is suggested, however, that it be not more than fceff = 0.9. In the case of a variable core design, it must be possible to preset these features to the calculated level before commencement of a particular experiment.

2.1.1.6. It must be possible to shut the critical assembly down under all circumstances. Normally, there must be two separate shut-down systems, each capable of shutting the assembly down. They may be of different design, including the operating mechanism to eliminate the possibility of similar simultaneous failures. Each system should be capable of absorbing 125 °/o of the maximum reactivity release. The time of operation of the shut-down devices, including detection and operating time before they are fully inserted, should be less than 50% of the reactor period when the maximum permissible single addition of reactivity is made above the critical value. One widely used method is free fall of absorbers from a position above the core and as near to it as possible. There may, of course, be other or secondary shut-down devices (e.g. a water-moderated assembly fitted With two systems of shut-down rods may have a moderator or reflector dump valve as a secondaiy shut-down device). The time of operation of such a secondary device need not be as short as that specified for the main or primary shut-down devices. Where the possibility exists that the system may become critical as a consequence of fuel addition during loading, some shut-down rods must be in the “ out” or suspended position during loading. For this purpose additional shut-down rods may be necessary. If so, a neutron flux above the lower level of adequate sensitivity of the detectors must be present and the rate of fuel loading should be such that the protective system is able to act with adequate rapidity.

2.1.1.7. The temperature coefficient of reactivity should also be investigated and the worst conditions should be estimated. If achoice is available, it is of course preferable to ensure that thiscoefficient be of negative sign, since this would assist the safetydevices in shutting down an assembly in the event of an accident.

2.1.1.8. The void coefficient of reactivity should similarly be investigated.2.1.2. A neutron source must always be used during an approach to

criticality, on starting up the assembly or when the shut-down

2 17


rods are in the “out” or suspended position, unless there is sufficient spontaneous neutron generation. The strength of the source must be enough to give a reading on the count-rate or flux-level meters. An interlock should be arranged to ensure the presence of this neutron source by means of a minimum value of count-rate or flux level reading. A mechanical interlock may be used to ensure that the source is in a safe position when the system has to be entered.

2.1.3. The design must be such as to ensure that the fuel is loaded insuch a way that no subsequent accidental movement can takeplace and cause a change of reactivity.

2.1.4. If he intends to consider the use in a critical assembly of fuel of varying degrees of enrichment or composition, the designer must ensure that for a particular lattice configuration only the correct fuel may be loaded into the assembly. This may be achieved by mechanical design of fuel rods and lattice plates so that only fuel of the correct enrichment or composition can be loaded into and retained in a particular lattice.

2.1.5. It may be intended to use a critical assembly as a zero-energyreactor after the initial approach to criticality has been carriedout with a particular fuel loading and lattice. If control rods are not required for the critical approach experiment, they may be incorporated into the assembly primarily for use at this stage. Suitable interlock and safety circuits must be employed to ensure that further fuel or moderator additions above the measured critical values are impossible when subsequent operation as a zero-energy system is required without the limitations in operation imposed upon critical assemblies. See Section 2.4.3.3. (Core Certificate, Section F), Section 2.4.6.5. (Operation of critical assembly — limitations) and Section 2.4.7. (Operation as zero- energy reactor).

2.1.6. The design of equipment and instrumentation important to the safety of a critical assembly should, where possible, be such that a failure of any component or piece of equipment which could result in dangerous conditions or non-detection of such conditions would cause the complete critical assembly to revert to safe shut-down conditions. This stipulation is based on the “ fail-safe” principle of design, whereby safety rods must be held out of the assembly so that a failure of any component would cause them to be driven in by some form of stored energy such as gravity; in liquid-moderated assemblies, it should be possible

18


to remove the moderator by gravity flow through a valve whose normal position is open; and a deliberate and positive action must be required to initiate or increase reactivity. This “fail-safe” principle must also be applied to all incoming supplies of electricity, compressed air or water; should any of these supplies fail at any time during the operation of a critical assembly, the assembly must automatically revert to a safe shut-down condition.

2.1.7. Shielding must be provided for all personnel in the vicinity of the assembly. It may be found desirable in practice to shield only the personnel connected with the operation of the assembly and to exclude all other personnel from the immediate area. The thickness of shielding and the extent of the exclusion area can only be arbitrarily assessed, since under normal conditions the radiation levels will be low. However, if the worst possible reactivity addition is assumed, the energy release which occurs before automatic shut-down by moderator or fuel dispersal may be calculated. Alternatively, the shield can be designed on the assumption that a power burst of 200 MW/sec will occur, this being the order of power releases which have, in fact, been produced in liquid-moderated critical assemblies before inherent limiting features shut the system down. This figure may not be applicable to other types of system. The consequent shielding thickness required may be calculated by assuming some limiting radiation dose to be received by the operators. The International Commission on Radiological Protection (ICRP) gives no definite recommendations on this subject but suggests that up to 25 rem occuring once in a lifetime shall be included as occupational exposure. Thus, 25 rem may be taken as the limiting dose to be received by the operators. It is important in designing the shield to ensure that personnel cannot enter the shield under any circumstances unless at least one bank of shut-down rods are in. A convenient design for the shielding may use portable concrete blocks of a size appropriate to the available lifting equipment.

2.1.8. Many other safety factors enter into the design of the assembly. Some of these are listed below. In seeking primarily nuclear safety, the designer must not overlook the important question of conventional safety. He must apply the codes of safe practice which are normally applicable to the design as regards such matters as mechanical hazards, guard procedures, pressure-vessel design and electrical hazards, to name but a few.

0* 19


2 .1.8 .1 .

2 .1 .8 .2 .

2 .1 .8 .3 .

2 .1 .8 .4 .

2 .1 .8 .5 .

2.1.8.6.

2 .1 .8 .7 .

2.1.8.8.

2 .1 .8 .9 .

2 .1 .8 .10 .

2 .1 .8 .11 .

Interlocking devices must be used to ensure the prevention of simultaneous additions of reactivity through independent actions. One shut-down system must be available during additions of reactivity if there is any possibility that the k^ of the system will exceed unity.If a neutron source is deemed indispensable during reactivity additions (see paragraph 2.1.2.), then the design should be such that reactivity additions are prevented when the source is not present.The structural materials used must be such that the expected corrosion or irradiation will not deteriorate them below safe limits of strength.Adequate ventilation must be assured, but provision must be made for it to be shut off if emergency conditions requiring this should arise.Interlocks must be so designed that their presence does not increase the hazards under emergency conditions.Shut-down rods must have a positive stop for their fully-in position. This stop should cause compressive rather than tensile stresses within materials, and should include some shock-absorbing system when this is necessary to prevent the rods from rebounding.The possibility that experimental thimbles or void tubes within the core of a liquid-moderated assembly may become flooded or float when the moderator level rises must not be overlooked.The possible consequences of an accident, and the measures which will have to be taken to restore or dispose of the equipment, must be studied. It may then be deemed advisable to incorporate special features which would simplify repair or other disposal. These features are specially important if they do not, in fact, give rise to undue expense in the construction of the system.The storage of fuel elements must be carefully considered. The prevention of accidental criticality must be a prime consideration and may be carried out both by geometrical design of the storage racks and by the use of neutron-absorbing materials. Attention must be paid to the possibility that the storage facilities may become flooded.It must not be forgotten that damge from external sources may constitute a hazard (e. g. fire, aircraft and earthquakes). The


2.1.9.

2 .1.10.

2 .2.

2 .2 . 1.

2.2.1.1.

shielding provided may in some cases provide sufficient protection against the first two of these hazards, but the third may call for special design features.An evaluation of the consequences of the failure of each component of the system must be made.When the design is completed, it must be critically reviewed to ensure that no hazards remain unprovided for, and also that no additional hazards have been created by some aspect of the design itself.

SAFETY IN INSTRUMENTATION OF CRITICAL ASSEMBLIES

GeneralEvery critical assembly must have equipment or instrumentation that will supply information on the physical parameters that describe and define the conditions in whioh the assembly is operating. Such equipment consists in a detector or sensor in the core of the assembly and an electronic channel to transmit the information to the operator. On the basis of this information, action to control the operating conditions of the assembly is taken, either at the judgement and discretion of the reactor operator or by previously determined and incorporated reactions of electronic circuits and other control devices. For example, a rise in temperature or flux level in the core of the assembly over a determined value would induce the operator to move control rods or would start an automatic action by electric circuitry to move control rods and shut down the reactor.The experiment to be performed on the assembly may call for information on parameters other than the basic parameters describing operating conditions. In some cases, additional instrumentation may be needed to provide this information, but more often the basic instrumentation supplying the working parameter data will adequately cover these additional needs as well. Where this is the case, the operations needed to obtain such additional experimental information from the basic control instrumentation must not be permitted to affect its proper functioning or to induce changes which are not compatible with the safe operation of the assembly. It should be remembered that the operator and experimenter may be two separate individuals looking at the assembly from two different points of view.

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2.2.1.3. The radiological protection requirements normally include permanent monitoring of the space around the assembly. Independent monitoring instrumentation supplying information on the level of radiation and radioactive contamination around the assembly is therefore needed.

2.2.2. Instrumentation needed

2.2.2.1. As a matter of principle, a critical assembly must never be operated unless two independent measurements of all parameters essential to the safety of the system are being made, i. e. two separate and independent instrument channels are working properly for each parameter.

2.2.2.2. The control instrumentation must cover every power range in which the assembly may be operating. This may require separate instrumentation channels supplementing each other’s range; two separate instrumentation channels covering at least two different ranges are necessary. It is further suggested that, as far as possible, the assembly should not be capable of operation, or should be shut down immediately, whenever any part of the basic control instrumentation is not working properly, by means of “fail-safe” features incorporated in the control instrumentation system rather than by operating instructions only. This requirement may result in many inadvertent shut-downs of the assembly, which may be extremely undesirable for the performance of the experiment. This situation can be avoided by adding some independent instrumentation channels (or some parts of them) and adding the principle of coincidence to that of duplication previously employed, (e. g., in a three-channel arrangement, signals from at least two channels would be required to cause reactor shut-down). This entails increased instrumentation costs. The cost of additional instrumentation must be balanced against the value of the operating time lost because of unnecessary shutdowns of the assembly when deciding the extent to which instrumentation should be duplicated. In some cases a similar effect to adding instrumentation channels can be achieved by overlapping ranges of different independent channels.

2.2.2.3. The positioning of detectors (or sensors) within the core must be carefully arranged to ensure that the information received corresponds to actual operating conditions of the assembly; e. g., neutron flux detectors must measure the neutron flux at a point where it is undisturbed by control rods so as to give information

2 2


on the operating power of the assembly. They must also measure the neutron flux of the core, not that generated directly by the source.

2.2.2.4. It is desirable that the electronic circuitry used be as simpleas possible; it may, however, be found useful in special casesto incorporate monitoring and test circuits which would greatly facilitate the checking procedures at start-up. These circuits are complex; they entail additional costs; being themselves liable to failure, they may increase the probability of unwanted shutdowns; and they are usually used only in connection with duplication and coincidence circuits (paragraph 2.2.2.2.). Their use in critical assemblies should therefore be extremely rare.

2.2.2.5. The radiological monitoring system for the space around the assembly must be fully as reliable as the control instrumentation. This is usually achieved by distributing a number of independent monitors at various points within that space.

2.2.2.6. Instrumentation for radiological protection should be designed to protect both operating and experimental personnel within the reactor building and personnel outside the building. Radiological monitors are required at vital positions inside the building; there must be monitors at any exhausts from the building. It is advantageous if the measuring instruments are set up at a pointseparate from the main control desk, so that this point may bemanned during operation by the radiological protection officer. Alarms indicating excessive readings should be placed at the control desk and at each of the measuring points. Permanent equipment to monitor the radiation level in various parts of the building must be installed, and it is usual to employ portable monitoring equipment in the immediate area of any special experiments.

2.2.2.7. Advice should be sought from the radiological protection officeron the provision of the detection equipment for use in emergencies.

2.2.3. “Fail-safe” design is an essential feature of the instrumentation and equipment used in critical assemblies. The channel should be tripped whenever equipment is disabled for any cause. A detailed list of such causes cannot be given in this Manual, as they depend very much on channel and instrumentation design, but the following are quoted as examples; loss of ion-chamber power supply, loss of power supply to instrumentation, open circuits in cables or instruments, short circuits (including those

23


2 .2 .3 .1 .

2.2.3.2.

2.2.3.3.

2.2.4.

2.2.5.2.2.5.1.

to earth) in cables or instruments. On failure, the reporting channel must initiate the same action as would occur if the highest permissible value had been exceeded. Depending on the nature of the channel, it should either give a warning signal to the operator or shut the assembly down immediately. An immediate shut-down must follow any failure of that part of the system between the point at which the independent channels join to cause control action and the control mechanism itself (e. g. control-rod drives).All control instrumentation should, as far as possible, be of “failsafe” design.Normally, “ fail-safe” design is not necessary for instrumentation used for experimental purposes, provided that proper operation of the control instrumentation is not affected.For the radiological protection monitoring system, “ fail-safe” design is recommended.Failures and consequent shut-downs because of power cuts may, if desired, be avoided in some cases by using an independent power supply for certain instrumentation. This may at the same time increase the precision of the measurements. Whenever a power cut might produce a dangerous situation, provision for an auxiliary power supply must be made. An independent power supply must always be used when a power cut would prevent the transmission of information essential to the safety of the assembly.

Control actions

When the information supplied by the assembly instrumentation indicates unsafe operating conditions, action must be taken to change those conditions and make the assembly safe. This action must not depend on the judgement and reactions of the operator if it affects the extent of an incidental power excursion. Usually such action is required in the shortest possible time; it must therefore be initiated automatically by the electronic circuitry of the assembly. For example, if the period meter indicates a period shorter than the preset value, or if it trips because of the failure of the period channel, it must cause an automatic shutdown of the assembly. Certain other parameters, e. g. flux or power level and temperatures, may call for similar arrangements. In some cases, a trip may be required if two independent warning signals (see 2.2.5.2.) have been received, e. g. an excessive

24


radiological monitor reading following by a warning of high exhaust activity. Safety trips must not be self-resetting. All trip arrangements should be critically examined to ensure that they are essential to safety.

2.2.5.2. There are some oases where a preset level can be exceeded without an assembly shut-down being necessarily required, and theautomatic action should then consist in warning the operator. This is the case, for example, of health monitoring systems. The warning signal should be both audible and visual, the latter also indicating the cause of the warning. It may be feasible to cut: off the audible warning without necessarily removing the cause, but it must not be possible to cancel the visual warning signals: except by removing the cause of the warning.

2.2.5.3. In certain instances it may be found useful to combine the twomethods given above for a vital parameter. In this case twopreset levels would be established: a level at which a pre-trip warning would be given and the trip level. The use of two preset levels may enable a trip to be avoided if the value of the parameter being measured is changing slowly enough for action by the operator to be possible.

2.2.6. Recorders should be used to keep a record of the critical parameters describing the behaviour of the assembly. The number of recorders should be sufficient to cover all the ranges of assembly operations.

2.2.7. Interlocks may be used in the electronic circuits connecting the signals received through the instrumentation to the control system to ensure the proper sequence of steps when any operation is being performed.

2.2.7.1. The use of interlocks in the instrumentation and control circuits must be carefully considered. They must be arranged to ensure safe operation but they must not cause any hazard under emergency conditions. It is impossible to be specific in recommendations on particular interlock requirements. Obvious cases where interlocks should be employed include those where potential hazards are materially reduced, where the operational routines are greatly simplified and where the additional cost is small. Interlocks should not be employed where their presence would demand the frequent use of cut-outs and where for small operational convenience great expense would be entailed.

2.2.7.2. One important form of interlock system is a line containing a

25


number of switches in series, all of which have to be closed before the required operation may continue or commence"'.

2.2.8. Interlock cut-outs may be necessary under certain special conditions of operation. Deliberate provision must be made for those in order not to encourage unauthorized cutting-out of interlocks by operating staff.

2.2.8.1. The need for cut-outs of interlocks must therefore be provided for. The use of interlock cut-outs must be very strictly regulated in order to avoid dangerous and uncontrollable situations.

2.2.8.2. In principle, cut-out devices must be built into the system (e. g. in the control panel) and their state must be visible at a glance. “Fail-safe” warning devices showing the cut-outs which are in use should be displayed on the control panel. It should be made impossible for the assembly to be put back into operation unless the cut-outs used previously have been removed. This can be ensured either by automatic removal of cut-outs on shut-down or by arranging that the assembly cannot be started up until the cut-outs are removed by positive action on the part of the operator.

2.2.8.3. Every precaution must be taken to ensure that the minimum number of cut-outs consistent with the basic principles of the reactor safety system are used.

2.2.8.4. One of the possible solutions of the cut-outs problem would be to ensure that:(a) Every out-out is connected to a keyed switch;(b) Keys are only to be drawn from the authorized holder by

authorized personnel;(c) For certain vital cut-outs two keys may be required, each of

which is held by a separate authorized holder;(d) Not more than a certain number of keys can be used at one

time. This number is one less than the minimum number required to cause dangerous conditions; and

(e) Keys are returned every day to their authorized holders.-2.2.9. The layout of the control panel or console must take into account

the limited human ability to perceive external sensations. The essential indicators, alarms and controls of the assembly must accordingly be concentrated in the central portion of the panel

* For example, each o f the reactor trips may be connected to the line controlling the safety rods in their withdrawn position.

26


2.3.2.3.1.

2.3.2.

2.3.3.

2.4.2.4.1.

2.4.2.

so as to be as convenient as possible for the operator. The less tation (e. g. that relating to the power supply) should be grouped around the central portion of the panel. Instrumentation and important indicators and warnings and any auxiliary instrumen- control equipment for experimental purposes which is not essential to the safety of the system should be displayed at a separate control position under the charge of a member of the experimental team. This ensures that the assembly operator is not distracted from his main task. The use of graphic panels in the design of control desks has proved most helpful. These graphic panels, in which switches and meters are incorporated, display schematically the system under consideration.

SAFETY OF CONSTRUCTION OF CRITICAL ASSEMBLIESThe usual safety codes regarding conventional safety must be adhered to. The applicability of these is obviously dependent upon the system under construction, and it is beyond the scope of this Manual to enter into the details of such codes.The components of the assembly forming the active core must be kept free of contamination by fissionable materials. Care must be taken that such components are adequately cleaned so that undesirable long-lived activation does not occur when the assembly is operated, and to prevent long-term corrosion effects.The inspection of the assembly and its components at all stages of manufacture must be of a high standard to ensure complete compliance with the design requirements. An adequate inspection service must be available for this purpose. The appropriate safety committee may require the results of the final inspection of the complete assembly before its clearance certificate can be issued.

SAFETY IN OPERATION OF CRITICAL ASSEMBLIESMuch of the information contained in this Section on staff organization and procedures, starting procedures and approach to criticality should appear in the safety report pertaining to the system or in its appendices.The staff organization must be clearly laid down and each mem- mer of the staff must not only understand his own duties but must also have a good knowledge of the duties of those senior and junior to him. In an experiment involving a large number of staff, there is a clear need for division of labour. A manager and, if necessary, a deputy manager for the assembly should be

27


2.4.2.1.

2.4.2.2.

2.4.2.3.

2.4.2.4.

2.4.2.5.

2.4.3.

appointed. Two groups of staff should report to the manager. These are (1) operators/experimenters, and (2) maintenance and ancillary staff. There must also be a radiological protection service available to provide the manager with guidance and advice in the field of radiation protection. This service will usually be provided by members of a radiological protection section or division within the establishment who are authorized to take independent decisions. In the case of an assembly involving a larger number of staff there will perhaps be a need for further subdivision of the staff.The manager ishould be in sole charge of the critical assembly at all times. All assembly staff report to the manager.The deputy manager, if one is necessary, should act as deputy to the manager in all respects. If shifts are being worked, there should be a deputy manager for each shift.Some of the operational and experimental staff may be responsible solely for the safe operation of the critical assembly, and would then act under the orders of the manager or deputy manager. There may be a chief operator who may have a staff of junior operators responsible to him. Other staff may be responsible solely for the conduct of experiments with, and in, the critical assembly. They must, however, obey the chief operator’s instructions when the safety of the critical assembly system is affected.The maintenance staff are responsible for the maintenance of the system, including periodic routine preventive maintenance. The ancillary staff include clerical staff, door-keepers, etc.The radiological protection service must advise the assembly personnel on all aspects of radiation safety, including personnel control (see paragraph 2.4.11.).The importance of adequate written procedures cannot be overstressed. These written procedures should form part of, or an appendix to, the safety report submitted to the appropriate safety committee. They should be supplemented by written instructions to each member of the operating team. Each member of the operating team must know the instructions and procedures applicable to other members of the team. Completely detailed procedures are rarely possible, since critical experiments are often designed to measure unknowns and they may produce unexpected developments. The written procedures and instructions referred to should therefore stress the principles involved and should note

28


especially those actions which may only be carried out by specified personnel, those actions which are forbidden, and the actions required in an emergency. Care must, however, be taken to avoid the suppression of initiative and intelligent action in unexpected situations. Members of the operating team must be given adequate time before the experiment is carried out to study these instructions and procedures thoroughly. They should be examined in these procedures before working on the assembly and, in addition, re-examined at regularly intervals and after long absences from the assembly. This examination, which may be carried out by the assembly manager, should cover, as well as normal operating procedures, the procedures necessary in an incident or emergency.

2.4.3.1. The authorized holder of the assembly keys may issue the keys only to authorized personnel. He may issue safety circuit cut-out keys only up to the number authorized and only to the manager or deputy manager (if they are not the holders).

2.4.3.2. It is important that an accurate and complete log book of all operations on the critical assembly be kept at all times. This book should be the responsibility of the senior man present, and he must ensure that all significant events are recorded.

2.4.3.3. The use of a system of core certificates has much to recommend it. This is a system which ensures that all personnel connected with the critical assembly operation know at any time the state of the assembly and the operations which may be performed. For assemblies with variable cores this certificate is essential. The certificate should contain spaces for the following items of information and should be displayed in a prominent position near the control dosk of the critical assembly. An example of the layout of a core certificate is given in Section 6.4.Section A of this certificate should contain a full description and a diagram of the particular core construction being employed at the time.Section B should contain a description of the control and shut- off rods. These two sections should be certified by the assembly manager and chief operator after the construction of the core has been completed.Section C should contain information regarding the estimated reactivity worth of the shut-down and control rods.Section D should contain estimates of the critical parameters of the assembly. Sections C and D should be signed and dated by

29


the physicist making these estimates and by the assembly manager. When the whole of this part of the lattice certificate has been completed, a critical approach may be made.Section E should contain spaces for the insertion of the measured values of the critical parameters and the measured values of the shut-down and control rod worths. These should also be signed by the assembly manager and the physicist after completion of the critical approach.Section F is optional, and is a certificate signed by two senior staff members that further fuel or moderator additions above the measured values given in section E are impossible. This may be ensured, for example, by the removal of ® key which governs the fuel or moderator addition system. The core certificate is now confirmed. It is at this stage that subsequent operation of the assembly as a zero-energy reactor may be permitted, if this is desired and if provision has been made for it in the initial design of the assembly by fitting control rods and by suitable interlock and safety circuitry.Section G of the core certificate should be a detachable slip. This slip, which is to be signed by the assembly manager and the chief operator, is the only form of order which is accepted by the personnel concerned and enables the critical parameters to be adjusted. It may be used to unload the reactor to permit core reconstruction or if additional fuel is required within the reactor system. In any case the removal and issue of this slip renders the core certificate invalid. A new certificate is therefore necessary before a fresh approach to criticality can be commenced.

2.4.4. The process of fuel handling may be divided into four stages: storage, handling from store to reactor, loading into the reactor, and unloading from the reactor into store.The organization and methods used for these stages must be such that accidental criticality does not occur. Storage has been considered in paragraph 2.1.8.10. The amount of fuel in transit at any time must be limited to avoid a dangerous- build-up of fuel at any point. An approved programme of fuel handling showing the fuel loading at each stage of the operation should be displayed. A fuel accounting system must be established to enable the receipts at the loading point to be checked against withdrawals from the store. The correct disposition of fuel in the reactor, in transit and in the store should be confirmed at each stage before the operation proceeds.

30


2.4.5. Procedures to be followed before start-up of the critical assembly must be compiled. These procedures should cover the tests which must be carried out on com pletion of manufacture and the tests and checks which must be carried out before each day’s operation.

2.4.5.1. After manufacture and installation has been com pleted all m echanical, electrical and electronic equipment must be thoroughly tested. Certain items o f mechanioal equipm ent may require endurance testing. W here appropriate, such test details as times of operation should be ascertained and checked for conformity with the specifications.

2.4.5.2. The issue of the clearance certificate by the safety committee may be dependent upon these tests and their results.

2.4.5.3. Before each day’s operation, tests and checks must be made toensure the correct functioning o f all equipment. The use o f thedetailed check lists by the operation staff is recom mended. These check lists should contain detailed instructions regarding the equipment to be tested and the test results that should beobtained. If there is a failure in any detail, the fault must beremedied before the operations are com m enced. These check lists must be signed after completion by the person carrying out the checks and certified by a senior officer.

2.4.5.4. Such a check list may contain, for example, detailed checks under the follow ing headings:

Reference to log books for special instructions;Maintenance com pleted;Incom ing services (electricity, air, water, etc.);Em ergency services;Supplies to individual instruments, meters and controls;Instruments functioning correctly;Instruments giving correct readings;Radiological protection instrumentation functioning;Com munication equipm ent functioning;Calibration checks of instruments;Safety circuits checked;Interlocks functioning correctly;Shut-off mechanism functioning correctly;Interlock cut-outs restored;Safety-limit switches preset to correct values;Variable range instruments on correct range;Valves and controls in correct position for start-up;Confirmation of the existence of accident instructions.

31


2.4.6. The first approach to critical with any new core. The generaloperating routines must have been laid dow n and presented in the safety report to the safety committee. The pre-start-up routines must be carried out. The keys for interlock cut-out may only be drawn against the signature of the reactor manager or his deputy. At least two persons must be present at all times. It is of vital importance that one person present should be in sole charge; all others, irrespective of rank, must be subordinate to him.

2.4.6.1. W here a chief operator has been nominated, he has sole charge of the assembly and must be in agreement with any orders issued by the manager through him.

2.4.6.2. The core certificate, if one is used, must be com pleted and signed. The log book must be present and a recorder appointed.

2.4.6.3. The first sub-critical core configuration may now be assembled. It is suggested that one-third of the predicted critical mass, if mass is the variable fundamental parameter, be the upper limit of the first core loading. If the variable parameter is not mass, the total fuel loading may be inserted, provided that the first addition of the variable parameter is limited to the equivalent of one-third of the predicted critical mass.

2.4.6.4. The source, if any, must be within the core and all instruments must be functioning correctly. The initial neutron level must be sufficient to be adequately detected. The shut-off rods may now be withdrawn. The routine governing their speed of withdrawal at the latter part of their stroke must be adhered to if it has been compiled, because criticality may be attained when the rods are withdrawn. T he approach-to-critical procedure may now commence.

2.4.6.5. In most cases the approach to criticality is made by increasing some fundamental parameter (e. g. fuel loading, moderator height) in successive amounts and taking the corresponding readings of the neutron levels (counting rate). If these additions cannot be made in sufficiently small increments or in a sufficiently accurately controlled way, a variable control parameter (e. g. moderator height, control rods) is necessary in order to make it impossible to reach criticality during the addition operations. Since it is important that a graph be kept of the inverse count rate or inverse flux against the fundamental parameter, in order that estimates of the critical value o f the fundamental parameter may be made, a sufficient number of incremental additions must be made to enable a good approach curve to be drawn. After each

32


addition of reactivity, a revised estimate of the critical value of the fundamental parameter is to be made. A safe rule is to add at every stage a maximum of one-third o f the expected value needed for criticality, up to a point very near criticality when the curve o f approach defines the critical point very clearly. At this stage the final addition may be m ade to achieve criticality. In every case at least two senior staff members (such as the manager and the chief experimenter or chief operator) must agree on the amount of each reactivity addition. After the limit suggested in 2.1.1.4. has been reached the increase o f the variable control parameter must be limited in speed to 2 X 10~20/o Ak/k per sec and must be interrupted at intervals as suggested in 2.1.1.2.

2.4.6.6. If there is a possibility that the system may becom e critical as a consequence o f fuel addition during loading, some of the shutdown rods must be in the “ out” or suspended position, a neutron flux above the low er level o f adequate sensitivity of the detectors must be present, and the rate o f fuel loading should be such that the protective system is able to act with adequate rapidity.

2.4.6.7. During the approach-to-critical experiment, estimates must be made of the worth o f shut-down and control devices in terms of the fundamental parameter which is being varied. The worth o f the control device, if any, is obviously the safe upper limit for the subsequent incremental additions of the fundamental parameter, but normally the addition must not be greater than 50°/o o f the worth o f the control device.

2.4.6.8. After criticality has been achieved and the experiment concluded, the assembly must be returned to a safe shut-down condition by inserting the shut-down rods and, if necessary, reducing the variable parameter. T he exact course of action will depend upon the design o f the assemply and upon whether a control rod is fitted. The lattice certificate should now be com pleted and confirmed.

2.4.7. If a control rod, or any other device which can regulate reactivity,is fitted, and if suitable provision has been made regarding the limitation o f the variable parameter (see Section 2.1.5.), the subsequent operation of the assembly as a zero-energy system after confirmation of the core certificate (Section F) may not be subject to the restrictions of paragraph 2.4.6., and may instead becom e

'3 33


2.4.8.

2.4.9.

2.4.10.

2.4.11.

2.4.12.

subject to the operational requirements given in Section 3 for Research Reactors.

If the conditions in paragraph 2.4.7. are not satisfied, subsequent approaches to critical and normal operation may be made by withdrawing the shut-down rods and increasing the variable parameter subject to the limitations in paragraph 2.4.6.

W here a possibility of hazard exists, fuel or moderator unloading must be carried out with at least two responsible personnel present.

Maintenance and checking schedules must be drawn up for the periodic maintenance o f the system and for preventive maintenance on a daily, weekly or monthly basis, and must cover all equipment, whether electrical, mechanical, or electronic. The assembly manager must be satisfied that the scheduled routine maintenance has been carried out before operations begin on any day. The perform ance of these maintenance tasks should be entered and certified in the critical assembly log book. The entries should indicate any faults that may have been found and their subsequent correction. Maintenance found necessary by the operating staff should be entered in the log book. This work must be com pleted before the system is re-started. Adequate time must be allowed for maintenance and repairs.

T o ensure that no changes are made during maintenance which will affect the safety o f the assembly, all maintenance should be authorized by the assembly manager by the use of a “ permit-to- work” system. This is particularly important in the case o f safety systems whereby a limit must be placed on the amount of equipment or absorbers which may be withdrawn at any time, and also where the actual work must be restricted to a given time period to ensure that an operation is not being carried out elsewhere on the assembly which requires for safety reasons, the presence of all the absorbers. The “ permit to work ' should include a statement by the radiological protection service regarding the health aspects of the work and prescribing in detail the precautions to be taken.

The risk of ingestion of radioactive dust or particles is reduced if smoking, eating, drinking and the application o f cosmetics are not permitted in the vicinity of critical assemblies. It is, therefore, recom mended that an establishment directive be issued to prohibit these activities in the places concerned.

34


2.5.

2.5.1.

2.5.2.

2.5.3.

2.5.4.

2.5.5.

2.5.6.

2.5.7.'

2.5.8.

3 *

The choice o f instruments to be used in the radiological protection work will to a considerable degree depend upon what other work besides that going on around the assembly is being perform ed in the establishment. Usually, however, certain general, basic instruments are required and the follow ing is an outline o f these.The individual exposure dose from external ionizing radiation may be measured by films or ionization chambers. The blackening of the films measured by a densitometer is almost independent of energy for energies higher than 250 KeV, while there is an increased sensitivity o f the films for energies below 250 KeV. Films have a special advantage in that they are well suited to be stored as records for later reference.T he neutron flux, whether fast or slow, may be measured by photographic emulsion. The tracks in the photographic emulsion are measured or counted with a microscope. The neutron emulsions are, however, easily fogged by storage, which may give rise to wrong readings, and in addition, the counting technique is quite difficult. The film used to detect slow neutrons is covered with cadmium foil.The dose may also be measured by means of small ionization chambers forming a condenser. They may be o f self-reading type or they may have to be read by a separate electrometer; they can be made either for gamma or neutron radiation.The film badges are considered superior to the pocket dosimeter or ionization chambers, mainly because they can be made to give a better indication o f the quality of radiation and form a permanent record.Facilities must be provided for monitoring persons leaving areas where radioactive contamination is possible. H and-and-foot monitors have been widely used for this purpose in larger establishments. Past experience has, however, shown that they require a good deal o f maintenance. G ood survey instruments are often sufficient to detect contamination on hands and feet below the established levels without necessarily using such expensive instalments as hand-and-foot monitors.The general beta-gamma survey instruments are based on Geiger- Miiller counters, ionization chambers or scintillation counters.

Survey instruments using Geiger-Miiller counters are easy for

RADIOLOGICAL PROTECTION INSTRUMENTATION

35


unskilled personnel to handle and seem to be perfected to such an extent that they may in most cases be used instead of ionization-chamber instruments. The Geiger-Miiller counter instruments may give incorrect readings in high radiation fields be cause of overloading of the counter. Special types exist which overcom e the energy dependence that limits the usefulness of the simple Geiger counter.

2.5.9. Slow neutrons are usually monitored by ionization chambers filled with boron trifluoride gas or coated with boron. The accuracy of the available instruments is satisfactory.

2.5.10. Instruments for fast-neutron monitoring are usually built on the principle of measuring the recoil protons produced in a hydrogenous material. Fast-neutron monitoring is still not advanced to a satisfactory state and involves difficulties. In particular, such commercially available instruments as do exist require frequent checking with a known neutron source.

2.5.11. Neutrons may also be monitored by foil-activation methods and the energy spectrum is determined by using different foils (threshold detectors).

2.5.12. Possible loose contamination may be detected by rubbing a piece o f filter paper over the contaminated surface on the spot in question and measuring the accumulated radioactivity on the paper (the “ smear-and-wipe” test).

2.5.13. Air monitoring is necessary in the evaluation of hazards from air-borne contamination. The air monitoring may be done by sucking air through an electrostatic precipitator or a filter paper, or by an impaction device. The deposited dust is afterwards measured for radioactivity. T he content o f natural radioactivity in the atmosphere is usually so high that the samples are left, after prompt evaluation, to decay before final conclusions are reached.

2.5.14. Continuous monitoring o f the air may be obtained by drawing air through a filter paper and simultaneously measuring the deposition by a detector connected to a rate-meter. Some elaborate air monitors with a moving filter paper are commercially available. The sensitivity o f the simplest type of instrument is, however, quite adequate and may in most cases even be superior to that of the moving-filter type.

2.5.15. As a general rule it can be said that the simplest instruments are usually the most reliable and easiest to maintain in working conditions. It is advisable to have two or more instruments of the same type so that inter-calibration may be carried out. One

36


should avoid instruments which need special types o f battery or components that are not readily available in normal research laboratories. If at all possible, instruments should be o f a type in which failure is directly indicated or results in high readings, Calibration sources or devices should be either built into the instruments or readily available at the place o f work.

37


3. R E S E A R C H R E A C T O R S

M uch o f the information contained in Section 2 on Critical Assemblies may be used in connection with research reactors. Obviously, however, not all o f it will be applicable and further information is required which is particular to research reactors. This additional information is contained in this Section. A danger which is inherent in a research reactor, but not normally in a critical assembly, is that of accidental release of fission products during an excursion, since usually the integrated flux will be m uch greater. In general it is assumed that the research reactors will be supplied for, rather than built by, establishments wishing to make use of this Manual. Section 3.1. on design and instrumentation is intended therefore as a guide to reactor specification. Clearly the owner o f such a reactor system has the final responsibility for its safety, and so must be responsible for the safety report. However, the owner would be well advised to require the supplier to make available to him sufficient information on the details and calculations o f the design and on the construction to enable him to prepare the safety report for presentation to the safety committee. T he owner should also require from the supplier some formal assurance that the safety of the design and construction has been adequately considered at all stages.

3.1. SAFETY O F D E SIG N A N D IN STR U M E N TA TIO N O F RESEARCH REACTO RS

3.1.1. Considerations of Reactors Physics3.1.1.1. It is desirable and possible in reactors of the small low -pow er type

under consideration to design the lattice so that a reactivity accident severe enough to rupture the fuel cladding could not take place. The total excess reactivity available within the system would then be less than that required to cause an accident o f that degree o f severity in any circumstance, including withdrawal of control rods, removal of the experimental absorber, flooding o f beam tubes and changes in the temperature of the system. If this approach were adopted, the likelihood of severe accidents affecting perhaps the surrounding area would be greatly reduced, since there could be no gross release o f fission products. The reactor should be so designed that more than one fuel element, or assembly, cannot be loaded at a time. Each fuel element or assembly should have at any position a smaller reactivity worth than the reactivity required to cause an accident

38


severe enough to rupture the fuel cladding, which means in effect that it is normally preferable to employ a system having a large number of small fuel elements or assemblies.

3.1.1.2. The reactivity invested in the control rods should be 50°/o greater than the total excess reactivity available within the core. This total excess reactivity must include that due to removal of the experimental absorber, experimental fissile sample insertion, flooding of beam tubes and alterations in temperature o f the system. The speed o f withdrawing the control rods should normally <be limited so that reactivity does not increase at a rate exceeding 2 X I 0 -2O/o kkjk per sec. The use o f an additional fine control rod with a lower speed of operation is also recom mended.

3.1.1.3. There should, in addition, be not less than two shut-down rods. Each o f these must be capable o f shutting dow n the reactor under all circumstances. It is important to ensure that criticality cannot be achieved on withdrawal of the shut-down rods.

3.1.2. In the mechanical design o f the reactor it is important to ensure that all vital equipment and features should be o f “ fail-safe” design, to the extent consistent with efficiency o f operation. It is preferable, but not essential, that observing equipment should be so designed.

3.1.2.1. In the overall design of the system it must be borne in mind that damage from external sources may constitute a hazard. The possibility of an external fire spreading to the reactor should be considered, although the biological shielding will usually impose an effective barrier. Another possible form of external accident is that of an aircraft crashing into the reactor itself. Again, the shielding may prove to be a considerable barrier against this hazard. In certain areas the possibility of earthquake must not be overlooked and if necessary special precautions should be taken in the design of the reactor and of its shielding.

3.1.2.2. It is important to ensure that adequate cooling o f the reactor system is achieved at all times. Any stoppage o f the coolant flow, or any loss o f coolant, if this is possible, must cause fast shutdown o f the reactor, and provision must be made for the heat generated by the decay o f the fission products to be adequately removed, so as to prevent melting o f the fuel elements. In many cases sufficient cooling may be achieved by radiation, by absorption or by convection within the coolant or moderator. Similar considerations regarding the decay heat apply if the

39


moderator or coolant is dumped on shut-down of the reactor.3.1.2.3. It is preferable to arrange reactor refuelling in such a way that

it is impossible to load more than one fuel element or assembly at any given time. In order to exclude any possibility of a subsequent movement of the fuel element into a region giving higher reactivity values, it should not be possible to load a fuel element partially, and the fuel elements must be firmly fixed in position to prevent their displacement by coolant flow or by vibration. The storage o f spare unused fuel elements must be carefully considered. Precautions must be taken against the accidental achievement o f criticality, against flooding and against unauthorized withdrawal of fuel.

3.1.2.4. Provision must be made for the removal o f fuel elements from the reactor and their subsequent storage. In the type o f reactor under consideration, it is unlikely that the fuel-element transfer flask would have to be cooled or that the fuel element store should be cooled, but this point should be investigated. In the design o f the store adequate precautions must be made also for the storage o f active wastes which may result from the reactor operation or from experiments perform ed with, or in, the reactor. These provisions must include the storage of these wastes, their subsequent disposal and, if necessary, their dilution or concentration. Active wastes may also arise as a result of fuel-element failure. In many reactors this contingency is now very remote, but it is necessary in all cases to make provision for it, primarily to ensure that the design does not hinder the operations that would be required; undue expense should not, however, be incurred initially in this respect.

3.1.2.5. The shielding provided around the reactor system must reduce the radiation levels to values recom m ended by the ICRP for the normal operation of the reactor. In reactors o f the type under consideration it is unlikely that this shielding would have to be cooled, but this point should also be investigated.

3.1.2.6. Provision must be made for rapid and easy mechanical maintenance o f the system.

3.1.3. The control instrumentation and interlocks of the system must be carefully designed. M uch of the information required may be adapted from Section 2.2. The use o f circuits em ploying the principle o f coincidence as well as that o f duplication, as described in Section 2.2.2.2., may be more profitable in the case o f research reactors. The extra expense involved does, however,

40


3.1.3.1.

3.1.3.2.

3.1.3.3.

3.1.4.

3.1.5.

3.1.6.

3.2.

3.2.1.

have to be balanced against the expense involved in unnecessary reactor shut-downs. The maintenance requirements o f the instrumentation system must also be considered.All readings of reactor parameters essential for the safety of the system must be duplicated. Readings in excess o f preset values must initiate either immediate reactor shut-down or warning signals to the operators, depending on their seriousness and upon reaction-time requirements. W here duplicated signals disagree in value, a warning signal should be displayed.The use o f interlocks in the instrumentation and control circuitry must be carefully considered. They must be arranged to ensure safe operation but they must not cause any hazard under emergency conditions. It is impossible to be specific in recom mendations on particular interlock requirements. Interlocks should obviously be em ployed in cases where potential hazards are reduced, where the operational routines are greatly simplified, and where the additional cost is small. Interlocks should not be em ployed where their presence would demand the frequent .use o f cut-outs and where great expense would be involved for a small return in operational convenience. It is also suggested that a keyed system of cut-outs be em ployed and that the cut-outs should be automatically restored on shut-down or alternatively that the reactor cannot be restarted until the cut-outs are restored.The state of the reactor must be displayed prominently in the reactor hall and in each experimental area. Reactor emergency shut-down buttons must be available at each experimental point, if necessary for safety.The radiological protection instrumentation requirements are noted in paragraph 2.2.2.6., to which reference should be made. An evaluation of the consequences of the failure of each com ponent o f the system must be made.On completion, the design must be critically reviewed to ensure that no hazards remain, and also that no additional hazards have been created by a portion o f the design itself.

CO M M ISSIO N IN G O F RESEARCH REACTO RS

The operations connected with commissioning are very like those described for the operation o f critical assemblies in Section 2.4. Paragraphs 2.4.5. and 2.4.6. particularly apply, although there may be differences in magnitude. Paragraph 2.4.4. should be

41


3.2.3.

3.2.4.

3.2.5.

3,3.

3.3.1.

3.3.1.1.

3.3.1.2.

noted in connection with fuel handling problems. Other features peculiar to the commissioning of research reactors are noted below.The tests of the reactor before the first approach to criticalitv are similar to those described in paragraphs 2.4.5.1. and 2.4.5.2., with the addition o f tests on the reactor cooling circuit. A summary of the results o f these tests must be forwarded to the safety committee by the reactor manager. The reactor manager and chief operator must be satisfied before com m encing the first approach to criticality that the tests have been carried out satisfactorily.The first approach to criticality may be carried out in a manner similar to the methods described in paragraphs 2.4.6. et seq. and demands the attendance of senior staff. Written procedures are required and these written procedures must have been submitted to the safety committee.After criticality has been achieved, all equipment must b s tested under operating conditions. Certain equipment cannot be fully tested under inactive conditions and must therefore receive special attention in operation.After com pletion of these tests, the reactor power-level may be raised in stages until the working value is achieved. The appropriate and necessary testing and checking at each stage must be com pleted before the next stage is proceeded with. This testing and checking is particularly necessary in connection with dynamic stability of the reactor, fuel-element temperatures (should these be measured), coolant temperatures, coolant circuits, and radiation levels within the reactor area; these can be com pletely checked only when the reactor is operating at power.

SAFETY IN O PERA TIO N O F RESEARCH REACTORS

The responsibilities and duties of the members of the staff of a research reactor must be clearly defined.

A reactor manager should be solely responsible for the research reactor. In his absence a deputy manager must be in sole charge.

The sections responsible directly to the reactor manager are operators, maintenance and ancillary staff. A chief operator may be appointed and have responsible to him the necessary number of shifts o f operators, each shift having a shift supervisor. In cases where no chief operator is appointed, the duties assigned

42


to him in this Manual would be undertaken by the reactor manager. The maintenance section is responsible, for- the m echanical, electrical and electronic maintenance of the equipment.

3.3.1.3. There may be a chief o f experimental services responsible to the reactor manager. H e is responsible for the experimental services of the reactor and may have a deputy together with a staff of experimentalists, maintenance personnel, a commissioning section to test equipm ent before it is put into the reactor, and ancillary staff. Experimenters using the reactor as a source o f radiation for special purposes will not belong to the staff o f the chief of experimental services. They should, however, be subject to his control when utilizing the reactor facilities, since this will ensure the co-ordination and liaison which are 'essential to the efficient programming and use of the reactor. The experimentalists must obey the orders of the operating staff during the operation of the reactor.

3.3.1.4. There must be a radiological protection service with authority to take independent decisions. It will have a chief and as many assistants as may be required, who are responsible for advising the reactor manager, the operating staff and the experimental staff on all questions concerning radiation hazards and for operating the health physics aspects of personnel control. This service will usually be provided by members of a radiological protection section or division within the establishment.

3.3.1.5. The reactor manager must report at frequent intervals to the operations safety committee for the particular reactor, as discussed in Section 5.7. In the case of a small reactor, this operations safety committee may also ensure the efficient programming and use o f the reactor, but this is in principle undesirable and it is better even in the smallest establishment not to com bine the two functions o f safety and efficiency. In all cases, therefore, the institution of a separate operations committee for the reactor should be considered. This operations committee should consist of all senior reactor staff. Its duties will include the programming of the experimental use of the reactor and of the operating schedules, etc.

3.3.2. Written procedures for operations and experiments must beissued after approval by a safety committee.

3.3.2.1. There must be detailed operating instructions for each memberof the operating and experimental teams. Each member of these teams must be aware of the duties of those around him.

43


The manager is responsible for ensuring that each member of the reactor staff is regularly examined in these procedures, especially those relating to emergencies, and particularly after long periods o f absence from the reactor.

3.3.2.2. Com plete records o f all reactor operations must be kept. Records must be kept of all important reactor parameters and these should be found useful in the routine preventive maintenance. Records must also be kept o f the experimental materials being irradiated, reactor fuel, both within and without the reactor, absorbers within the reactor, health physics information, including radiation surveys, and active effluent quantities. Modifications to the system or its operations must be recorded. Such m odifications may require the approval o f the safety committee if the safety of the system is affected. L og books describing experimental operations must also be kept if these operations relate to reactor safety.

3.3.2.3. Any deviations from established procedures and any experiments or irradiations may be made only with the written approval of the reactor manager. This approval will authorize the operating staff to permit such deviations, experiments or irradiations.

3.3.3. General procedures for operation3.3.3.1. At normal reactor start-up, the shift supervisor and operating

staff and the appropriate radiological protection staff must be present. A start-up check list similar to that in 2.4.5.4. must be used. Any fault found must be remedied before the start-up is permitted. Estimates must be made o f control-rod positions for criticality. The start-up routines may then be com m enced. After major modifications or refuelling, the chief operator must also be present.

3.3.3.2. For normal shut-down of the reactor, the shift supervisov, operating staff and the appropriate radiological protection staff must be present. The reactor may then be shut dow n by the agreed procedures. A check list must be used to ensure the com plete safety o f the system. This check list must be signed by the shift supervisor. It should contain spaces for shut-off rod positions, control-rod positions, incom ing services disconnected (except those which are required for safe shut-down reactor conditions, for example continued fuel-element cooling), all keys returned, together with other specific requirements.

3.3.3.3. During normal operation of the reactor, at least two persons must be present at all times. If continuous operation is used,

44


3.3.3.4.

3.3.3.5.

3.3.3.6.

3.3.3.7.

3.3.4.

3.3.4.1.

a check list should be completed at the commencement o f each shift. This check list should contain the relevant portions o f that given in Section 2.4.5.4. The safety circuits should be checked and if possible monitored and tested at the beginning of each shift. If the coincidence-duplication principle has been employed in the safety circuits, the faults in the safety equipment must be immediately corrected in the faulty channel. Other faults must be corrected as soon as possible. If an unexplained shutdow n occurs, the reactor must only be re-started after exhaustive investigations and with the chief operator’s consent. If other shut-downs occur, the reactor may not be re-started unless the fault which caused the shut-down has been remedied. Records o f the reactor operation must be kept as in Section 3.3.2.If a key system for interlock cut-outs has been adopted, the keys may only be m ade available to authorized persons and in accordance with the conditions laid dow n by the reactor manager. At the conclusion oi operations, tests of safety circuits, including the shut-down rod actuating mechanisms, ought to be made at periodic intervals by injecting the appropriate signals into the safety circuits or by reducing the trip levels. All trip functions should be regularly tested in this manner.The radiological protection service must ensure that the areas around the reactor and experiments are being monitored. It must also advise the operational and experimental staffs o f any special precautions to be taken at any time. It must also advise the reactor operating staff o f the need to shut dow n the reactor when undue radiation hazards make this course desirable. The risk of ingestion o f radioactive dust or particles is reduced if smoking, eating, drinking and the application o f cosmetics are not permitted in the vicinity o f the reactor or in any other area where the operations may produce a radioactive contamination hazard. It is therefore recom mended that an establishment directive be issued to prohibit these activities in the places concerned. Every experiment carried out at the reactor must have its own chief responsible to the reactor manager in so far as safety is concerned. Experimental staff must obey the orders o f the operating staff when these are connected with safety or with the agreed operating procedures.W hen the safety of the reactor system may be affected by faults in experimental equipment, the equipm ent must be connected to the reactor safety circuits to effect reactor shut-down. Annun

45


ciators should indicate to the reactor operator the condition of this experimental equipment.The reactor operating staff are responsible for ensuring that the various experimental requirements do not conflict, and for the efficient programming o f operations for the reactor. If a separate committee is form ed as suggested in Section 3.3.1.5. it will be found very valuable for this purpose. Care must be taken to ensure that experimenters operating such equipment have sufficient space to avoid mutual interference.The reactor operating staff are to accept materials for irradiation only with the written permission of the reactor manager and must ensure that no material which is hazardous or unsuitable for irradiation is inserted into the reactor. The reactor operating staff must have adequate control o f the irradiation facilities to ensure that such materials are not introduced into the reactor. This applies especially to pneumatic irradiation services, whose loading point may b e remote from the reactor hall. Strong absorbers or fuel samples so loaded are potentially dangerous because of the high speed o f loading and unloading. The reactor operating staff must also ensure that experiments containing strong absorbers are not removed from the reactor without their permission.Maintenance and refuellingThe preventive maintenance o f all equipment must be on a scheduled basis. The reactor manager must ensure that such preventive maintenance is being carried out, especially before start-up of the reactor system.Corrective maintenance of safety equipment must not be carried out unless the reactor is shut down, except where the equipment incorporates the principle of coincidence-duplication and the reactor is fully protected during the whole period.To ensure that no changes are made during maintenance which will affect the safety o f the reactor, all maintenance should be authorized by the reactor manager by the use of a “ permit-to- work” system. This is particularly important on safety systems where a limit must be placed on the amount of equipment or absorbers which may be withdrawn at any time, and also where the actual work must be restricted to a given period to ensure that an operation is not being carried out elsewhere on the reactor which requires the presence o f all absorbers for safety reasons. The “ permit-to-work” should include a radiological pro-


tection service statement on the health aspects o f the work and details o f the precautions to be taken.

3.3.5.4. Refuelling the reactor system must be recognized as a potentiallyhazardous operation. A strict routine must be established to ensurethe use of only the correct grade of enrichment of fuel. (See alsoparagraph 2.4.4.) The procedures for the whole operation must bewritten and the operation may only be carried out with the chief operator present. This procedure may require only fuel assembly to be changed at a time.

3.3.5.5. After refuelling, reactor start-up is a potentially hazardous operation and should be treated as such.

3.3.5.6. A flask for the transport of irradiated fuel elements may b e necessary. Storage for irradiated fuel elements must be provided and this storage should preferably be external to the reactor system if defective fuel elements may have to be stored before disposal.

3.4. E X PE R IM E N TA L USE O F RESEARCH REACTO RS

3.4.1. In the design of experiments and design and manufacture ofexperimental equipment many factors must be taken into conflux-measuring devices o f the presence of such equipment, the succeeding sections.

3.4.1 1. The main considerations to be taken into account in such designwork are nuclear interactions, reactivity effects of loading and unloading and failure of equipment, the effects on the flux or flux-measuring devices of the presence o f tsuch equipment, the heating due both to neutron and gamma-radiation, which may be experienced in such equipment, the shielding requirements o f equipm ent and the custody and handling o f fissile samples.

3.4.1.2. The control and instrumentation of such experiments must be adequate, and where the safety o f the reactor may be affected such instrumentation must be duplicated. It is also advisable for interlocks to be arranged between the equipm ent and the reactor circuits where the safety of the reactor itself may be affected.

3.4.1.3. Chemical considerations in the design of such equipment include compatibility (for example, the explosion hazards on failure of sodium-containing equipment immersed in water), the com position of materials (especially organic materials) under irradiation, radiation-induced corrosion and the consequent blocking of cooling channels, mass-transfer effects and the suitability of cleaning" and degreasing materials.

4 7


3.4.1.4. Metallurgical considerations in the design of such equipment include radiation damage to structural metals, the compatibility of metals and materials, thermal effects, induced radioactivity (for example, the cobalt content o f materials should be minimized if at all possible because o f the long-lived cobalt-60 produced under irradiation) the dimensional change o f fissile samples and corrosion effects. Special attention must be paid to brazing methods and the fluxes used.

3.4.1.5. The engineering design and manufacture must be to acceptable and nationally recognized safety standards. If remote handling o f the equipment is considered, special attention to design must

‘ be given to enable this to be easily carried out. The materials used should be appropriate for conditions of irradiation. The effect of the failure of high-pressure circuits must be considered and protection fitted if necessary. Pressure-release devices should not cause any hazard in the reactor or the reactor building. Emergency cooling o f equipment, especially fissile samples, may be required. The effect o f mechanical failure of moving and rotating machinery must be investigated and protection fitted.

3.4.1.6. It is recom m ended that extra precautions should be taken in experiments containing fissile material. These precautions may take the form o f (installing fission-product detectors or o f providing additional containment.

■3.4.1.7. Experimental equipm ent must receive adequate preventive routine maintenance.

3.4.1.8. All equipment to be loaded into the reactor must be thoroughly cleaned to prevent the contamination o f the reactor and the formation of active contaminants under irradiation.

3.4.1.9. Thorough inactive testing o f all equipment under actual operating conditions must take place before loading such equipment into the reactor. The reactor manager will require the results of such tests before giving permission for the equipm ent to be loaded into the reactor.

3.4.1.10. The incom ing supplies o f electricity, gas and air available at the reactor must be adequate for the operation o f the experimental equipment.

3.4.1.11. All experiments using open beams must be adequately shielded to reduce scattered radiation to a safe level. Emergent beams must be stopped by “ beam catchers” (or traps). It is the responsibility of the radiological protection personnel to ensure that the

48


3.4.2.1.

3.4.2.

3.4.2.2.

3.4.3.

3.4.3.1.

3.4.3.2.

3.4.3.3.

3.4.3.4.

3.4.3.5.

above requirements are complied with before the reactor is started up.

T he experimenter is responsible for producing the safety report for the experiment that he wishes to carry out.

W ritten safety reports must be submitted at the various stages to the appropriate operations safety committee described in Section 5 b y the experimenter concerned through the reactor manager. These reports should deal with general procedures, including details of liaison between operating and experimental staff. Final submission to the committee may have to be made after the conclusion o f the commissioning tests if these have been required.The reactor manager may obtain the committee’s agreement to specified routine irradiations and experiments without individual prior reference to the committee. In these circumstances he may issue his approval personally but he must submit reports of experiments so carried out to the committee at frequent intervals.In the operation of experimental equipment a primary requirement is that no hazard be occasioned to personnel or to the reactor.The experimental staff must be responsible for the safety of their own equipment.

It must not be forgotten that failure of the equipment, even though it may not represent a hazard to the reactor, may present considerable hazards in the subsequent unloading operations because o f the activities involved, and this may cause costly delays to the rest o f the reactor programmes.

Experimental equipment must be operated only with the knowledge o f the appropriate operations safety committee and the approval o f the reactor manager. The operating staff must have given their approval and must be informed immediately of the start-up and shut-down of such equipment.

Care must be taken to ensure that only experimental materials approved by the reactor manager are loaded into the reactor.

As an undetected failure o f experimental equipment may cause release of radioactive materials, special care must be taken when removing equipment from the reactor to ensure adequate protection against radiation and contamination.

49


3.4.3.6. The reactor personnel must be aware of the state of any experiment, perhaps by means o f annunciators in the control room. Any interlock circuits connecting the experiment to the reactor safety system must be regularly checked and must at all times be working.

3.4.4. The liaison necessary between the experimental and operating staff must be carefully thought out, especially when the safety of the reactor is in question. The experimental staff must obey the orders of the reactor operating staff for loading, unloading, or commencement o f operation of experimental equipment. They must keep the operating staff inform ed o f the state and progress of the experiment. They must immediately inform the operating staff o f failure o f any of the experimental equipment in order that such action as may be required to minimize any potential hazard may be taken by the operational staff. The operating staff have general responsibility for the control and safety of the whole experimental area adjacent to the reactor.

3.4.5. Measurements may be made at a point remote from the experimental apparatus which is itself set up at an experimental channel o f the reactor. I f provision has been made for opening and closing this experimental channel from the measuring station, careful liaison is necessary between the experimental staff and the radiological protection staff, who should be inform ed in advance of each operation affecting the experimental channel, as other personnel may be affected.

3.5. R A D IO L O G IC A L PR O T E C T IO N IN STR U M E N TA TIO N

3.5.1. Paragraphs 2.5.1. to 2.5.15. are directly applicable to researchreactors.

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4. P E R S O N N E L

4.1.

4.1.1.

4.1.2.

4.1.3.

4.2.

4.2.1.

It is o f vital importance that all personnel engaged in critical assembly or research reactor operation should have the necessary qualifications and experience and have received training appropriate to the duties undertaken.

In such operations all staff must, at all times, act in strict accordance with any instructions and regulations which are issued. One o f the primary objects of the training period should be the inculcation of discipline in these matters. It is known that under conditions of routine operation disregard both of the hazards and o f obligatory regulations can occur. T o ensure that personnel are always ready and watchful for the onset of an emergency is very difficult. Accidents occur very rarely, and this fact tends to cause com placency among operating personnel and the supervisory staff.

It is recom m ended in Section 4.2. that a nucleus o f senior staff members should have received prior training in reactor technology, and if possible, some experience o f the operation o f research reactors or critical assemblies. Attendance at a recognized reactor course is most desirable for these senior members. W here no facilities exist in the country concerned, it is possible for such training to be taken by prospective senior staff members in other States, and arrangements for such training may be made with the aid o f bilateral or multilateral agreements or o f international or regional organizations. If a research reactor is being supplied, it is most desirable that the supplier should afford senior staff members an opportunity to take a period o f orientation and familiarization training on a similar reactor elsewhere before the reactor is delivered. If a portion of such a training period could be spent on actual operation, its value would be greatly enhanced.

Q U A LIFIC A TIO N S

Minimum health requirements should be established for each position in the staff organization. This is necessary to ensure correct perform ance of the duties assigned to each member of the staff. Such health requirements should include minimum standards o f sensory perception and standards concerning strength, stamina and age. Emotional stability and personal character are also o f great importance. If a possibility exists of ex-

INTRODUCTION

4 * 51


posure to radiation, the staff mem ber’s previous radiation exposure history should be considered.

4.2.2. Formal education requirements are difficult to define in this Manual because o f differences in practice in various M ember States. Managers and deputy managers should have received a university or equivalent level o f training in engineering or science and it is desirable also that at least one of them should have attended a recognized reactor course. The chief operators, operators, and chief o f experimental services of a research’ reactor should have received not less than a high school or equivalent education, preferably with a scientific or technical bias. The formal education requirements of experimental staff can be considered only in connection with the duties they have to perform.

4.2.3. Some indication of desirable minimum requirements of previousexperience are given below. The manager or deputy manager and the chief operator of a critical assembly or research reactorshould have had a minimum of three months’ practical experience in the operation and use of a research reactor or critical assembly. In the case o f research reactors which have been supplied from elsewhere they should also have attended a period of initial training (and if possible, operation) on a similar system (see Section 4.1.3.). M embers o f the maintenance section, especially the electronics maintenance staff, should, if possible, have received a short training period on a similar installation to enable them to becom e familiar with the equipment.

4.3. ON -SITE T R A IN IN G

4.3.1. In a certain sense the training of all staff .should be a continuousprocess, especially in the matter of radiation safety, emergency procedures, accident case studies, and discipline, as has been noted in Section 4.1.2.

4.3.2. H owever, for operating staff (operators, shift supervisors of research reactors, ancillary staff) a short course o f training which should be mainly practical in nature is most useful. Experimental staff may also benefit from the relevant sections o f such a course. This training should include relevant items from the list below :(a) Orientation period on the particular installation.(b) Sufficient instruction on reactor physics and reactor theory

to give staff an adequate understanding o f their practical duties and enable them to take intelligent action in an emergency.

52


(c) Instruction in health physics, the effects of radiation, radiation safety practices, and the personal prevention of over-exposure.

(d) For operators: a period o f supervised operation, follow ed by supervised start-up and shut-down procedures and fuel-changing procedures.

(e) Instruction in the emergency procedures in effect at the particular installation.

4.3.3. It may be desired to organize a more com plete reactor training course. As such a course entails the use of high-quality lecturing staff and extensive experimental facilities, it should be organized only where the number of potential students warrants it. It may be organized on a regional basis with the assistance o f international or regional organizations. As an example o f a comprehensive course, a syllabus with 95 hours of lectures is given in Appendix III. W ith experimental work, the duration of this course would be two months.

4.4. E X A M IN A T IO N A N D C O N TR O L

4.4.1. On com pletion o f the on-site formal training, the trainees should be examined in both theory and practice. They must pass this examination successfully before they are permitted to take up the duties of any particular post.

4.4.2. Regular periodical reviews, with special reference to the subject matter o f their duties and to radiation safety and emergency procedures, should be conducted for all staff. These periodical reviews may be held yearly; such a review should also be held after a staff member has been absent from this duties for a long period.

4.4.4. The supervision of staff qualifications and internal examinations may be carried out by an establishment committee appointed by the Director for this purpose. In a small establishment where no other alternative is possible, the Safety Committee may be made so responsible. As is pointed out m Section 5.1.6. this latter course is undesirable, since the Safety Committee will then have a regulatory function whereas its main function is purely advisory.

53


5. A D M I N I S T R A T I V E P R O C E D U R E S

The executive chain of com m and must at all times be responsible for the-safety of a reactor or critical assembly, and nothing must be done to diminish this responsibility either directly or by inference. As in all potentially hazardous occupations, the responsibility of each individual must be clearly defined and unambiguous; the definitions should be written in order to avoid misunderstandings; and it is an obvious principle that a man must be responsible to one man only above him in the chain of com mand if conflicting orders or situations are to be avoided.

5.1. T H E SAFETY C O M M IT TE E SYSTEM

5.1.1. It is standard practice in many fields of work to submit resultsor proposals for independent scrutiny, in order that an independent judgement may be obtained. This is an obvious necessity in the field o f safety and particularly in the field of nuclear safety, where the independent advisory body is the Safety Com mittee. In the case of nuclear installations it must be made ob ligatory, by direct order of the executive chain of command, to secure such advice. The use to which the advice so obtained is put also rests with the executive chain o f com m and; it has, however, been found in practice that the recommendations so obtained are always immediately and directly adopted.

5.1.2. The advice must be sought at many stages of the process of building and utilizing a nuclear installation. The conception, design, construction, inactive testing, commissioning and utilization of a nuclear installation are all stages at which advice should be obtained. After the final advice before actual utilization has been obtained, it should be made obligatory to secure further advice whenever significant alterations affecting the safety of the installation are proposed. Such alterations include those in which the built-in reactivity of the system is to be increased by any means; in which the safety circuits of the reactor are to be altered in any manner; in which the operational procedures are to be changed; and in which any hazardous materials are to be placed within the reactor system, unless advice has been previously sought on any o f these topics. This list is not com prehensive and is only indicative of the type of alteration that should require advice from the safety committee. It is further suggested that such advice must be sought at specific times at intervals not exceeding one year, to cover the possibility that a

54


series o f minor modifications may, taken together, constitute an important modification in respect of which advice should be obtained.

5.1.3. A dvice may be obtained from a single person, but since the safety o f the system depends upon many and diverse disciplines and technologies, it is recom m ended that it be obtained from groups of people skilled in these matters. It is highly desirable that these groups o f people should be currently practising, and readily conversant with, the various skills they represent, and thus safety committees com posed of part-time members are recommended.

5.1.4. The work required from these committees is, firstly, to advise on the overall safety of com plete reactors and critical assemblies and, secondly, to carry out more detailed appraisals of proposals concerning the operation and experimental use o f these installations. An establishment safety committee should be form ed to deal with the former, while the more detailed work may be carried out by an operations safety committee. It may be convenient to form a separate operations safety committee for each research reactor or critical assembly. There might be a measure of common membership between committees. The duties and membership o f these committees are described in the succeeding sections of this Manual.

5.1.5. The appointment o f members of the establishment safety com mittee must be subject to the approval o f the D irector o f the establishment, to whom the Chairman o f the committee must report. The appointment of the members of operations safety committees should be subject to the approval of the establishment safety committee to whom they will report.

5.1.6. It is important to note that the function of the committees is advisory and not regulatory. Primarily they advise the Director on the safety o f reactors and experiments within the establishment. In practice they also advise proposers and users on safety matters, thus providing an opportunity for plans and operations to be revised so that the committee’s reports to the Director will be satisfactory. It is best if the committees have no regulatory functions at all, but where, in a small establishment, it is essential to include among the duties such regulatory functions as programming, training and investigations of over-exposures, great care must be taken to ensure a strict demarcation of its duties. It is vital that there should be no diminution of responsibility for safety on the part o f the executive chain o f command.

55


5.2.

5.2.1.

5.2.2.

5.2.3.

5.2.4.

5.3.

5.3.1.

5.3.2.

Proposals for sub-critical assemblies are to be placed before the committee, which will advise whether or not they fall within the definition o f sub-critical assemblies given in Section 1.4.1. If they do fall within this definition, the committee will issue a certificate to this effect enabling work to proceed without any further reference to the committee.In the case o f critical assemblies and research reactors the safety reports must be examined by the committee with a view to assessing the safety o f the design and manufacture of the system. The operational procedures must also be examined, and the staff organization and proposals for selection of staff must be seen. The proposed procedures for emergency measures must be reviewed in the light of the best information available. Reports on the results of these examinations must be submitted by the committee to the Director o f the establishment.The committee will also receive reports, perhaps two or three times a year, from each of its operations safety committees. These reports must be studied to ensure that the safety o f operation and use of the particular assembly or reactor has been satisfactory. If necessary, comments or instructions must be forwarded to the appropriate operations safety committee.If no other committee within the establishment is available, the committee may also be made responsible for drawing up policies and standards regarding training levels and the experience required o f all reactor and critical assembly staff. W here this is the case, the operations safety committee o f each assembly or reactor would be responsible for advising the reactor or assembly manager on these topics, and would report to the safety committee.

CO M PO SITIO N O F T H E SAFETY C O M M IT T E E

T he committee members should be appointed on a part-time basis only. In large establishments, the volum e of work may be sufficient to justify a full-time secretary o f the committee, since the burden of work on the secretary is bound to be great. The members should be currently practising their arts and skills, and should be of some seniority.It is advisable that the committee, as a whole, be unconnected with and disinterested in the proposal on which advice is to be given. This is not meant to im ply that a minority of members may not be connected with the proposal under discussion, but

DUTIES OF THE SAFETY COMMITTEE

56


that the advice given by the committee as a whole should b e impartial.

5.3.3. It is recom m ended that the minimum size o f the safety com m ittee be four, the follow ing skills being represented:

Experimental Reactor Physics;Theoretical Reactor Physics;Health Physics;Reactor Operations and Technology; andInstrumentation and Control Engineering.

W here a larger volum e o f work, or the problems involved, call for wider representation, the committee may be increased in size up to a maximum number o f ten members. The follow ing additional skills may be found extremely useful on such a committee:

Engineering Design;Reactor Metallurgy;Reactor Chemistry; andChemical Engineering.

5.3.4. It is preferable that the Chairman of the committee, w ho reports directly to the Director of the establishment, should be otherwise unconnected with the design or operation o f critical assembly or reactors.

5.3.5. A representative of the site emergency organization may be present at meetings of the committee to ensure the necessary liaison between the project and the emergency organization.

5.3.6. There should, in addition, be a secretary o f the committee. The duties o f the secretary include the tendering o f advice to a proposer on the contents of the safety documents to be submitted and their form. H e must thus be conversant with the safety requirements of the establishment and the arrangements for liaison with the establishment emergency organizations. H e should be given sufficient official time for this work.

5.3.7. In a very small establishment, difficulty may be found in obtaining a sufficient number of independent members to form a committee. In this case recourse must be taken to one o f the follow ing three alternatives. Firstly, it may be found possible to recruit some committee members from other establishments within the same country. Secondly, other countries may be willing, perhaps on a reciprocal basis, to provide some part-time committee members. Thirdly, the help of international organizations could be sought. In the last case, part-time members may be made available from the staff o f the organization, or the organization may be able to

5 7


5 .4 .

.5.4.1.

5 .3 .8 .

.5.4.2.

5.4.3.

5.4.4.

58

assist in arranging for them to be made available from other countries.Even in large establishments, the use of some part-time committee members from other countries or establishments is very desirable in order that diverse experience may be called upon. The use of such members may also be helpful in ensuring the gradual unification of safety standards.

SAFETY C O M M IT T E E PRO CE D U RE

T o enable the safety committee to function correctly, and to enable it to give considered judgement and advice, it is suggested that proposers be obliged, by directive of the head of the establishment, to submit written safety reports to the appropriate safety committee for its consideration. These reports, in a sufficient number of copies for all members of the committee, should be in draft form. T he committee should be enabled to give its advice at the earliest stages. After the deliberations of the com mittee on these documents, the final report, incorporating the suggestions made by the committee and adopted, should be issued for general distribution to all interested persons. W here draft safety reports have been presented in stages, the final safety report issued for distribution should contain all the information presented at the various times. T h e committee will authorize for inclusion in the report a form of clearance certificate which will enable the proposer, without further reference to the D irector of the establishment, to proceed.The operations safety committee for each reactor or assembly must report to the establishment safety committee at specified times, say two or three times a year. The safety committee must therefore meet at least as often, but during the conception, design and construction of a new system it will also meet as required. The safety committee must be informed immediately of any major incidents ocurring with or in a reactor or assembly.It has been noted previously that the clearance certificate is to be of limited validity, say one year, after which time a re-submission may be made to the committee, showing what modifications have been made to the assembly or reactor, its m ode of operation, and its present and projected use, since the last submission to the committee.It is to be noted that the issue of the clearance certificate does not in any way diminish the responsibility for the safety of the


5.5.

5.5.1.

5.5.2.

5.5.3.

5.5.4.

system of the reactor or assembly manager. The clearance certificate is merely to state that in the opinion of the committee the reactor, assembly or experiment described in the safety report is safe, provided that the use to which it is put and the manner o f operation are identical with those prescribed in the safety report.

D U TIES O F TH E O PERATIO N S SAFETY C O M M ITT E E

The committee has as its main concern the safety standards of all operations and experiments with or in the particular critical assembly or research reactor to which it is attached. Modifications to assembly or reactor systems which affect the safety o f the system are also included within its scope, but the committee may desire (or the establishment safety committee may request) that major modifications be considered by the safety committee.

The operations safety committee should examine all proposals for use, modifications and experiments with or in critical assemblies and research reactors. In the case of large experiments requiring much equipment, advice may be sought from the committee at the conceptual, design, manufacture and pre-testing stages. In the case of small experiments, advice may be required at the final stage only.The committee should be inform ed at each meeting of any incidents which may have occurred during the operation o f the assembly or reactor that may have some relationship to the safety of the system.A secondary function of the committee may be to ensure the efficient use and programming of the assembly or reactor. In addition, if there is no other committee available for the purpose, the committee may include among its duties the investigation of cases o f over-exposure to ionizing radiations of personnel working with or around the reactor or assembly. This aspect o f its work is, however, primarily w’ithin the field of health physics. The committee may also advise the reactor or assembly manager on the requirements as regards training and experience o f operational and experimental staff, and may formulate in detail the requirements and policy laid dow n by the safety committee (Section 5.2.4). These aspects of its work are not considered further in this Manual, since it is not recom m ended that a committee which is primarily advisory in character should undertake regulatory duties (Section 5.1.6.).

59


5.6.1.

5.6.2.

5.6.3.

5.6.4.

5.6.5.

5.6.6.

5.6.

The committee members should be appointed on a part-time basis only. The members should be currently practising their arts and skills, and although they may be junior in rank to the members of the safety committee to whom they report, they should have had adequate experience in their respective fields.The principle o f independent judgement should be adhered to, and the majority o f members should be unconnected with the proposals under discussion.The minimum size of the committee should again be four. The reactor or assembly manager should be one of the members, and the following skills should be represented:

Experimental Reactor Physics;Theoretical Reactor Physics;Reactor Operations and Technology;Health Physics; andInstrumentation and Control Engineering.

In addition other disciplines, as suggested in 5.3.3., may be incorporated.The chairman of the committee should be independent of the reactor or assembly in question.A representative o f the site emergency organization may be invited to attend meetings o f the committee when matters affecting that organization are being discussed.

The comments in Section 5.3.6. also apply to the secretary of the operations safety committee. T o ensure smooth operation o f the committee system, there must be adequate liaison between the secretary o f the establishment safety committee and the secretaries of the operations safety committees. In a very small establishment, this ist best ensured by arranging for the same person to hold all the secretarial offices. In a very large establishment, requiring more than one person for these posts, secretaries may be drawn from a safety organization within the establishment.It is expected that even in the smallest establishment no external membership of operations safety committees would be necessary. This is obviously desirable, even at the expense o f a partial sacrifice o f the principle o f independent judgement, since the committees will meet very frequently.

COMPOSITION OF THE OPERATIONS SAFETYCOMMITTEE

60


5.7. O PERATIO N S SAFETY C O M M IT T E E PRO CE D U RE

5.7.1. It is expected- that the operations safety committee w ould meet very frequently, say once or twice per month, particularly if it is also to be concerned with the duties suggested in paragraph5.5.4.

5.7.2. T o enable the committee to function correctly and to give considered judgement and advice, it is recom m ended that intending critical-assembly users and proposers of experiments on or with reactors be obliged, by directive of the head of the establishment, to submit written proposals to the appropriate committee. Advice will be given by the committee on the basis of these reports.

5.7.3. T o enable routine experiments and irradiations to be carried out without the delay caused by individual reference to the committee, the reactor manager may request from the Director, by means of a written report, permission to carry out certain routine experiments and irradiation in research reactors without prior reference to the committee. His request, and the subsequent agreement, must be precise and within well-defined limits. Subsequent requests by reactor users, which may be made to the manager on a simple form, provided that they fall within these limits, may then be authorized directly by the reactor manager. A suitable form for routine irradiations is given in Section 6.2. W here this form is used, the committee should be notified at specific intervals by the reactor manager o f the routine experiments and irradiations that he has authorized since his last report.

5.7.4. In the case of simple experiments, use may be made o f a form(Section 6.3.) copies of which, when com pleted, are circulated to members of the committee. The committee is then enabled to give its advice.

5.7.5. In the cases o f alterations to the system or o f large experiments,a full safety report may be required, draft copies of which shouldbe circulated to all committee members. After deliberations of the committee on these documents, the final report, incorporating the suggestions made by the committee and adopted, should be issued for general distribution to the establishment safety committee and all interested people. The committee will authorize for inclusion in the report a form o f clearance certificate which will enable the proposer to proceed.

5.7.6. The clearance certificate is to be o f limited validity, say one year, after which time a re-submission must be made to the committee showing what modifications have been made to the experiment,

61


its mode o f operation and its present and projected use since the last submission to the committee. It is to be noted that the issue o f the clearance certificate does not in any way diminish the responsibility for safety of the experimenter to whom it is issued. The clearance certificate is merely to state that, in the opinion of the committee, the alteration or experiment described in the safety report is safe, provided that the use to which it is put and the manner o f operation are identical with those prescribed in the safety report. The reactor manager, being finally responsible for overall safety, has the right to refuse to allow the experiment to proceed.

5.7.7. The reactor or assembly manager must report on the operation of the system to the committee at each meeting and particularly on any incidents which may have occurred and which may bear some relationship to the safety of the system.

62


6. S A F E T Y D O C U M E N T S

6.1. SAFETY REPORTS

6.1.1. Safety reports to be presented to the committees have tw o main purposes. Firstly, they enable the committee to advise on the safety o f the system. Secondly, they cause the assembly, reactor or experimental staff to consider the safety of the system very carefully. It has been found from experience that it is at the stage o f writing draft reports that most o f the modifications necessary for safety are determined.

6.1.2. T w o types o f safety report are in general initially required. The first type of report concerns a new reactor or critical assembly. In the case of a reactor or critical assembly which is supplied by another organization, a design safety report should be made available. This design safety report should contain sufficient information for the final safety report to be prepared by the reactor manager. For reactors and critical assemblies designed and manufactured under local control, it may be advantageous for the design information to be presented to the safety committee in the form of interim reports at two or three stages, as described in 6.1.5. below. The second concerns experiments to be performed with or in a critical assembly or research reactor. This will be presented to the operations safety committee o f the particular system, and may be in one, two or three stages according to the nature and size o f the project. For convenience, both types of report have been considered together in paragraphs 6.1.5. to 6.1.10.

6.1.3. T w o other types of safety report may also be required. The first concerns re-submission to the safety committees before the expiration o f the clearance certificate (Section 5.1.2.), if it is intended that operation should be continued beyond that date. This report should refer to the initial report, and should contain details o f alterations required to any section o f the initial report as a result of modifications which have been carried out. The other report which may be required concerns modifications to' the equipment, its m ode o f operation or its experimental use w hich it is intended to carry out. This report should refer to the initial report, the modifications required and the consequent amendments necessary in sections o f the initial report.

6.1.4. The author of the safety report must be the critical assembly chief, the research reactor manager or the experimenter, in the

63


case o f critical assemblies, research reactors and experiments respectively. H e may be assisted by such staff as are necessary in the preparation. Assistance may be required from the designers or operators. If a reactor is supplied, the owner should require from the supplier sufficient information to enable a safety report to be prepared (Section 3).

6.1.5. In the case o f locally controlled projects, the committee will bein a position to provide advice at the earliest stages o f a project — thus permitting better safety practices and preventing waste in re-design or modification o f equipment — if interim safety reports are presented initially at the conceptual stage of projects. In small projects, the report at this time may contain com plete information, thus enabling the committee to give its clearance certificate without further reference to the committee. In the case o f larger projects sufficient information may not be available at this stage and further interim reports may have to be put before the committee. The information presented to the committee at each stage should be the maximum available, enabling work to proceed in the knowledge that the project is regarded by the committee as basically sound. The final published report containing the committee clearance certificate should contain all the information, revised where necessary as the result of agreed modifications presented to the committee in the various interim reports.

6.1.6. Safety reports are also required when alterations as described inSection 5.1.2. are contemplated, and at the expiiy o f the validity of a clearance certificate.

6.1.7. Safety reports should be clear and unambiguous and shouldenable the committee to give its advice without excessive recourse to verbal interpolation at the committee meetings. They should, on the other hand, be as concise and brief aspossible consistent with the foregoing desiderata. They shouldbe illustrated when necessary, preferably with simple drawings and sketches rather than reduced-scale working drawings.

6.1.8. The final safety report for critical assemblies and research reactors may be presented in two parts. The first part, which is mainly concerned with the design of the system, may have the follow ing sections:I. (1) Identification, description, purpose and programme.I. (2) Safety principles.I. (3) Description of control and safety systems.

64


I. (4) Monitoring equipment.I. (5) Equipment failure studies.

The second part, which is mainly concerned with the operation of the system, may have the following sections:II. (1) Commissioning tests and start-up procedures.II. (2) Normal operating procedures.II. (3) Administrative control.II. (4) Maintenance procedures.II. (5) Emergency procedures.II. (6) Miscellaneous information.It is suggested that the report could usefully reflect the subjects discussed in the previous sections of this Manual. The above section headings are amplified in the succeeding two paragraphs. The lists given are reasonably comprehensive and certain items may not be applicable in all cases.

6.1.9. The first part of the safety report, that concerned with design, should discuss the following items.

6.1.9.1. Identification, description, purpose and programme.(a) The identification of the project, together with a sufficiently

detailed description of it to enable the remainder of the report to be followed.

(b) The purpose of the project, together with the programme it is proposed to follow.

6.1.9.2. Safety principles.(a) The safety principles which have been adopted in the design

and operation of the project. The relative importance of mechanical and electrical automatic shut-down devices and administrative control should be shown, together with the degree of supervisory control by senior staff which is necessary.

6.1.9.3. Description of control and safety systems.(a) Maximum rates of reactivity additions, and maximum quan

tities possible by normal operation of each of the control devices, individually. Since these values (in conjunction with items in the next section) govern the negative reactivity safety devices, they should be kept as small as possible consistent with reasonable start-up times and experimental necessity.

(b) The maximum gain in reactivity expected to arise spontaneously or through accident of any kind (e. g. breakage of

5 65


control mechanism, temperature variations, displacement of fuel, etc.) and the rates of gain.

(c) The negative reactivity, and rates of application, controlled by the shut-down devices, showing the excess which has been allowed for failure of a proportion of the shut-down devices themselves.

(d) Method of determination and estimates of the coefficients of reactivity, in respect of temperature, coolant or moderator voidage and power.

(e) Shielding provided and expected levels of radiation external to shielding under conditions of normal operation, and failure conditions.

(f) A full and detailed description of all safety equipment, especially from the operational standpoint. Manual control mechanisms, automatic trips, interlock devices, interlock cutouts, warning devices and continuous motoring and recording devices should be included.

6.1.9.4. Monitoring equipment.(a) Radiation monitoring equipment provided, and routine pro

cedures for health physics purposes.6.1.9.5. Equipment failure studies.

(a) An appraisal of the effect of failure of equipment or personnel, and combinations of these failures. An assessment of the resultant consequences is necessary, together with a statement of the hazards so created. This will necessitate a study of the potential excess reactivity present, the reactor dynamics (including temperatures likely to be reached), the possible dispersion of radioactive material, the damage caused to the reactor and the possibility of fire.

6.1.10. The second part of the safety report, that concerned with operation, should discuss the following items.

6.1.10.1. Commissioning tests and initial start-up.(a) Tests of the system and components before fuel loading.(b) Initial fuel-loading and critical-approach procedures.(c) Tests of system and components after criticality has been

achieved and during the process of bringing a reactor to full power.

6.1.10.2. Normal operating procedures.(a) Operating procedures, particularly those concerning the nor

mal start-up, normal shut-down and special experimental use.

66


(b) The control of fissile material to the reactor, storage of fresh fissile material, and storage and disposal of used fissile material. If materials of differing enrichments are to be used, the measures taken to ensure their complete identification and automatic prevention of error should be stated.

(c) Fuel loading and unloading procedures.(d) The control of neutron absorbers and reflectors in and around

the reactor (other than those built into the reactor).(e) Active effluent disposal and results of a radiation survey of

environment before commencing operations.(f) The control of flammable material.

6.1.10.3. Administrative control.(a) The staff structure chart of all operational and experimental

staff. A detailed statement of the duties of each member of the group is also necessary, as are the qualifications required of each member. The names of the personnel undertaking each of these duties should also be added, probably as an appendix. (It is important that this appendix be kept up to date by re-submission to the committee as necessary. It is equally important that this appendix includes the names of deputies for the various described posts to cover unavoidable absences through sickness, etc.)

(b) The standing orders will be issued to each member of the team and to storekeepers, maintenance operatives and doorkeepers. (These must also be kept up to date by reference to the committee when necessary, and therefore their issue as appendices should be considered.)

6.1.10.4. Maintenance procedures.(a) The maintenance arrangements which will be made.

6.1.10.5. Emergency procedures.(a) The proposed arrangements to deal with emergencies which

may arise. Fire, explosion, spillage or dispersal of radioactive material within the building, and the dispersal of active material external to the building are included under this heading. The liaison necessary with the establishment emergency organization, health physics section (including decontamination service), fire section and medical section should be carefully explained.

6.1.10.6. Miscellaneous information.(a) For assemblies having variable core arrangements, an example

67


of the core certificate which is to be filled in at the time of completion of the core by responsible members of the team and before operation (see Section 2.4.3.3.).

(b) Probable requirements regarding future submissions to the project and a statement regarding future submissions to the committee for the project.

(c) Any other relevant information.6.1.11. Safety reports concerning experiments and experimental equip

ment should discuss in addition to the relevant topics of paragraphs 6.1.8. to 6.1.10., the appropriate items of the additional list noted below.(a) Mechanical loading on, and interference with, components of

the reactor system.(b) Heat-transfer and fluid-fiow considerations, including heat-

generation due to irradiation.(c) The necessity for filters in the flow streams, gaseous or liquid,

in the apparatus.(d) Statement of any radiolytic effects.(e) Discussion of any possible corrosion effects.(f) Stored energy in any part of the experimental system, e. g.

Wigner energy in crystalline substances.(g) Unusual requirements of reactor operation.(h) Test programme before starting the experiment, periodic tests

during the experiment and the test programme after removal of the experiment from the reactor.

6.2. FORM FOR ROUTINE IRRADIATIONS6.2.1. If the reactor manager has been so authorized, he may give his

permission for routine irradiations within a research reactor (Section 5.7.3.).

6.2.2. An example of a form suitable for this purpose is given below.

68


(Quadruplicate Form) R e q u e s t f o r r o u t i n e i r r a d i a t i o n

T o: Reactor ManagerFrom : .................................

Date

M aterial to be Irradiated,

Composition: Neutron Absorption Cross-Section: Physical State: W eight of Sample: Container Provided:

IrradiationR equired

Flux-Level:Time:Number o f Irradiation: Special Requirements regarding location in Reactor, and cooling:

IrradiatedSam ple

Expected Activity Level:Place of Use: Transport M ethod: Use:

D ate R equired bij:

From: Reactor Manager DateT o: Operating Staff (C opy 1, 2 & 3); Requester (Copy 4)

The irradiation has been authorized. Details

Modifications required

From: Operating Staff Date ............... ........................T o: Reactor Manager (Copy 1); Requester (Copy 3); (Copy 2 in Operating

Staff files)The irradiation has been carried out. Details ........ ........................

Measured activity level ..................................................................... ........................

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6.3. FORM FOR SIMPLE EXPERIMENTS

6.3.1. Where an experiment is of a simple nature it may be found expedient to prepare a form for intending experimenters, copies of which should be distributed to the Operations Safety Committee (Section 5.7.4.).

6.3.2. An example of such a form is given below.

(Sufficient copies for each member of the Operations Safety Committee)

P r o p o s e d e x p e r im e n t

To: M ember of Operations Safety Committee Date .......................................

From: ......................................................................................

T he experiment described below is proposed. A review by the Operations Safety Committee is requested.

Purpose:Experiment

Equipment to be used: Location:Safety Principles Adopted:

Personnel responsible: Proposed Date o f Experiment:

Additional Comments:

6.4. CORE CERTIFICATE

6.4.1. In operating critical assemblies, particularly these with variablelattices, the use of core certificates has been recommended (Section 2.4.3.3.).

6.4.2. An example of a core certificate follows. In this particular casea light-water-moderated assembly using uranium enriched in U835 has been taken as an example for the detail of the certificate. Since no control rods have been fitted, it is unlikely that this assembly would be operated as a research reactor at zero power, and Section F would not, in fact, be required.

70


E x a m p l e o n ly

Number: C ore.

T h is c e r t i f i c a t e is i n v a l i d if s e c t i o n “ g ” is n o t p r e s e n t S ig n a t u r e s

S e c t i o n A — C o r e c o n s t r u c t io n

1. Lattice used ............. (Dia- 4. T op cover locked on? ......gram) ..........................

2. Total fuel present in core 5. Total moderator quantity......kg present in system: ..... liters

3. Fuel type and enrichment or ..............cm height in re-.......................... actor tank

S e c t i o n B C o n t r o l a n d s h u t - d o w n r o d s

6. Construction o f shut-downrods ..........................

7. Position o f system 1 withinlattice ..........................

8. Position o f system 2 withinlattice ..........................

9. Above moderator height of.............cm ; pumping ratelimited to .............liters/seco r ............ cm /sec in reactortank

10. Above this height, singleadditions limited t o .............liters, o r .............cm heightin reactor tank

S e c t i o n C — E s t im a t e s o f c o n t r o l r o d a n d s h u t - d o w n r o d w o r t h s

11. At keif = 1, worth o f system 1 shut-down rods

12. At keli — 1, worth o f system 2 shut-down rods .

AkITAfcIT

S e c t i o n D — E s t i m a t e s o f c r i t i c a l p a r a m e t e r s

13. Estimated value of moderator height for feeff = 1, ......cm14. Estimated value of moderator height for feeff = 0.9, ......cm

A k15. Estimated value of excess ----- when moderator height isk °

equal to (5 ) ..........................16. Estimated value of moderator height increase for addition

of 0.5 %> Ak~ k

2 X 10—20/o sec .............cm /sec17. Estimated value o f rate o f height increase for addition o f

A k~ k

18. Is (11) greater than (15) X 1.25?19. Is (12) greater than (15) X 1.25?20. Is (14) greater than (9)?21. Is (16) greater than (10)?22. Is (17) greater than (9)?

Manager

Date

ChiefOperator

Date

Manager

Date

Physicist

Date

A p p r o a c h - t o - c r i t i c a l e x p e r i m e n t m u s t n o t b e m a d e u n l e s s a l l d e t a i l s h a v e b e e n c o m p l e t e d a n d a l l a n s w e r s a r e “ y e s ” .

71


E x a m p l e o n ly

C e r t i f i c a t e

C r it i c a l a s s e m b l y

Date of Issue..........................Date ofWithdrawal ..........................

S e c t i o n E — M e a s u r e d v a l u e s

Safety rod worths system 1 ............ , System 2 ............ PhysicistHeight o f moderator fceff = 0 .9 ............

fceff = 1 ............cmcm

o/ Ak u/o —— k

ManagerModerator height increase equivalent to O.r

cm Date

S e c t i o n F — C e r t i f i c a t e (if required)

Action has been taken as follows:(e. g. Removal o f key controlling moderator-addition system to ensure that the value of the critical parameter cannot exceed those given in Section E.)Assembly may now be operated as zero-energy reactor.

Manager

Tim e and Date Chief Operator

(Detachable slip) Certificate N o .........................

S e c t i o n G — A u t h o r i z a t i o n

Time and DateT o : .....................................( Section).

Critical Assembly

You are authorized to carry out the following:(e. g. Removal of top cover to permit alteration o f core)

Manager

Chief Operator

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6.5.6.5.1.

6.5.2.

6.5.3.

The reactor or assembly manager should report at frequent intervals to the operations safety committee on the operations carried out with the reactor or assembly (Section 5.8.7.). His report should consider in some detail any incidents which have occurred, and which may have had some relationship to the safety of the system.If the reactor manager has been personally authorized to approve irradiations within the reactor (Section 5.8.3.) he should report at frequent intervals to the operations safety committee on the irradiations so approved.Examples of start-up check lists, maintenance records and log book, have not been included. Such examples may be found, if required, in the published literature, notably the first reference in Appendix I, Section A.

OTHER DOCUMENTATION

73


7.1 .

7 .1 .1.

7 .1 .2 .

7.1.2.1.

7.1.2.2.

7 .1 .2 .3 .

7 .1 .3.

7. E M E R G E N C Y P R O C E D U R E S

Reference should be made to the IAEA Manual on the Safe Handling of Radioisotopes"' for amplification of many of the points covered in this section.

GENERALThe occurrence of a hazardous incident as a consequence of the operation of a reactor or critical assembly is highly unlikely if proper operational procedures and clear lines of authority are worked out before the start-up and frequently reviewed. In spite of this, plans covering every foreseeable incident should be made, in order to ensure that the organization can successfully deal with all consequences that might conceivably follow from such an incident.The extent and consequences of the incident will classify it within one of the following three definitions:Local incident.The local incident is confined to one building or part of it, and means any deviation from the foreseen operational conditions involving radioactive materials which may imply a hazard for personnel or equipment.Site emergency.The site emergency means any incident involving escape of radioactive materials from the local area or the presence of ionizing radiations that may constitute a hazard to personnel or property within the boundaries of the establishment.Public emergency.The public emergency is any set of circumstances involving radioactive materials which may cause hazards or damage to persons or property outside the boundaries of the establishment. The local incident definition covers most of the incidents which are liable to happen. This type of incident will involve only a limited number of personnel. It will most likely be confined to one room or a small part of a building. The local incident may grow as a result of the release of radioactive materials into, or the presence of ionizing radiation in, the surroundings. If the consequences are still confined to the area under the control of the establishment, e. g. inside the fence, this incident comes under the category of a site emergency. This site emergency may deve

Safe Handling of Radioisotopes, IAEA, Vienna, 1958 (STI/PUB/1).


7.1.4.

7.1.5.

7.1.6.

7.1.7.

7.2.

7 .2 .].

lop, releasing or threatening to release undesirable amounts of radioactive materials to the surroundings outside the confines of the establishment, and it is then called a public emergency.It is evident that the three types of incidents call for different kinds of planning. The public emergency will not only involve the establishment and its personnel but may require protective measures to be taken by part of the surrounding population: restrictions on the use of foodstuff produced in the area; restrictions on living habits, for instance small children being kept indoors, etc. Such plans and their execution must be prepared in co-operation with the Government authorities and other organs involved, such as the local police, fire department, civil defence and the health authorities.The site emergency will probably not call for any outside action, but the responsible persons and authorities concerned with the public emergency should be notified that a public emergency may arise and be alert. Similarly, the man in charge of any local incident should inform the responsible persons that a site emergency may arise.Any plans to deal with an emergency should as far as possible be periodically rehearsed, with particular regard to the procedures of notification, issue of warnings, and instruction of personnel in evacuation routes and use of protective clothing.It should be noted that the definitions given above include incidents which involve contamination, radiation or a combination of both. The emergency plans should consider each of these aspects.

LOCAL INCIDENTThe instructions regarding normal working procedures must contain the necessary general instructions for handling conceivable local incidents designating the responsible officer among the personnel present at any time, describing the types of warning signal and their meanings and indicating the lines of communication to obtain the necessary help for health physics, medical, decontamination and other necessary services. The administration should be made aware that parts of the establishment might in certain circumstances have to be closed off.The instruction regarding normal working procedures should contain the necessary instructions to deal with the most urgent tasks during a local incident, such as reduction of reactor power, immediate shielding against radiation, evacuation of personnel, warn

75


7.2.4.

7.2.5.

7.2.6.

7.2.7.

7.2.8.

76

ing personnel in adjacent areas, stopping ventilation, prevention of entry of unauthorized persons, roping off contaminated areas, use of respirators if necessary, and further instructions to prevent the local incident from growing.Where the incident involves radiation, only the points detailed below are applicable:(a) The radiation source at the place of the accident should, if

possible, be returned to its correct position, where adequate protection is afforded against radiation;

(b) During treatment of the incident, excessive exposure should be prevented by such methods as shielding, distance and limitation of working time;

(c) Attention should be paid to the protection of persons in the adjacent area, including rooms above and below. Areas subject to high radiation levels should be clearly marked and, if necessary, roped off.

There should be a record of the amount of radioactive materials present in the building and the extent of a release to the surroundings which would constitute a site emergency.If the local incident is associated with conditions in which an escape of radioactive materials from the local area exceeding what is believed to be a safe level might occur, the site emergency warning must be given. It may also be desirable to call a site emergency when the local incident is of such a nature that site emergency provisions are of value in meeting the local situation.An up-to-date list should be kept of those buildings where a local incident could cause a site emergency.Most of the local incidents will probably involve some type of contamination and will therefore call for subsequent decontamination measures. The health physics section is responsible for controlling these operations, but the following necessary procedures are listed for information:(a) Discovery and assessment of the contamination;(b) Limitation of contamination;(c) Reduction of contamination; and(d) Verification of the efficacy of decontamination.The decontamination of work places and equipment should be carried out in such a way as to prevent the contamination from spreading further, preferably by wet methods, or by other suit


7.2.9.

7.2.10.

7.2.11.

7.2.12.

7.3.

7.3.1.

7.3.2.

7.3.3.

7.3.4.

7.3.5.

able methods, such as those using adhesive tapes or stripping paints, that cause the least practicable contamination of the air during the cleaning operations and the smallest practicable spread of contamination to other areas.Spilled radioactive liquid should be absorbed by means of suitable material, such as blotting paper, sawdust or flannel.If radioactive powder is spilled, all fans and ventilating systems likely to raise dust should be stopped immediately.Decontamination should be carried out by the smallest practicable number of workers equipped with appropriate clothing.All cleaning appliances should be:(a) Used solely for cleaning;(b) Decontaminated after use; and(c) Kept in suitable receptacles.

SITE EMERGENCYThe organization to deal with a site emergency may be faced with many different tasks, such as warning site personnel, evacuating personnel from specific buildings or from the whole area, controlling traffic, monitoring the area, issuing protective equipment to workers, providing decontamination and monitoring facilities, alerting key personnel for the emergency provision of communications and overall supervision.During a site emergency a Site Emergency Officer should be designated to be in charge of the operations on behalf of the Director and he should be considered as the senior officer at the site.A duty rota of Site Emergency Officers should be maintained. A Deputy Site Emergency Officer should be appointed among the shift personnel to be in charge of the operations until the Site Emergency Officer arrives. If no technical staff are on site at night one senior officer should always be on call.The Site Emergency Officer must use all available resources to ensure the safety of the site as a whole, and to render any necessary assistance at the scene of the incident that has caused the site emergency. He must be assisted by the necessary staff to relieve him of the detailed work and assume control under his directions when required.A site emergency team should be appointed to ensure that the

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7.3.6.

7.3.7.7.3.7.1.

7 .3.7.2.

7.3.7.3.

7.3.7.4.

7 .3.8.

7.3.9.

Site Emergency Officer has sufficient staff at his disposal. Members of this crew should also be on duty rota.The site emergency team may consist of representatives from the administration, the engineering service, the medical service (if any) and the health physics service.The responsibilities of the different services may be: Administrative services:

Co-ordination of transport, police, fire and telephone, evacuation of personnel;Welfare, supply of food and drinks;Liaison with authorities outside the establishment; Declaration of a public emergency as required; and Arrangements for supply of extra personnel.

Engineering services:Supply and maintenance of emergency equipment;Site decontamination and waste disposal;Putting up road barriers; andSupply of personnel for various engineering tasks.

Medical service:Treatment of casualties; andArranging for treatment of casualties at hospitals.

Health Physics service:Supply and maintenance of health physics instruments; Advising personnel on health matters;Monitoring and decontamination of personnel;Monitoring of surroundings;Instruction in and supervision of the use of protective equipment; andEvaluation of the need to declare a public emergency. ,

One man should be appointed to make records containing all such information as time of arrival of information, decisions taken, action completed and persons involved.The warning must be given by the most appropriate means (sirens, klaxons or over loudspeakers). The personnel must be aware of the action they are to take on hearing the site emergency warning. The site emergency team, under the Site Emergency Officer, must be responsible for giving subsequent instructions to the personnel. The Site Emergency Officer will give notice to those buildings where evacuation is necessary or where normal work can be started again.

78


7.3.10. The equipment at monitoring stations for personnel should include:(a) Paper and pencil;(b) Radiation detection instruments with range of about

0.1— 1 mr/hr to 100 mr/hr;(c) Detergents and brushes for decontamination of people. Water

is presumed to be present. If electricity is available, vacuum cleaners with filters are useful;

(d) Filter paper for smear tests;(e) Plastic bags and labels for wrapping of contaminated mate

rials; and(f) Protective clothing and breathing masks.

7.4. PUBLIC EMERGENCY7.4.1. The public emergency will not only involve the establishment

and its personnel but may require protective measures by part of the surrounding population. Such plans and their possible execution must be prepared in co-operation with the Government authorities and other organs involved, such as the local police, fire department, civil defence and the health authorities.

7.4.2. If the site emergency develops into a public emergency, the right to make decisions concerning its control will pass to the public authorities. In practice the site emergency staff will continue to be responsible for implementing the necessary control actions under two circumstances:(a) Until such time as the appropriate public officials are in

a position to take over the direction of the emergency, the site emergency staff should in any event be prepared to deal with all public aspects of the incident which could develop in the first 24 hours;

(b) In the event that the establishment’s staff in part represents the Government’s pool of experts in the field of atomic energy, it can be expected that elements of the public authority will be vested in them. It should be clear in such circumstances that they represent the Government and are no longer acting as officers of the installation.

7.4.3. The public emergency staff should in principle consist of the site emergency staff with the addition of responsible representatives from the local police, the local authorities, the civil defence and a medical officer with atomic energy experience. Specialists

79'


in legal matters and public relations may be of great value. It can be expected that the overall direction of the emergency staff will be co-ordinated by an officer provided by and responsible solely to the central government authority.

7.4.4. The present Manual does not deal in detail with procedures to be adopted, but consideration must be given to the following items:

Evaluation of the hazard;Notifying the population;Supplying information to Government officials;Control of radiation sources and contamination;Protective measures for the affected population;Monitoring of subsequent radiation levels;Complication of records; and Practice of emergency procedures.

7.5. REQUIREMENTS OF THE HEALTH PHYSICS SERVICE

7.5.1. The health physics service is essential for the determination of the hazards caused by an incident and for its subsequent control.

7.5.2. In addition to the normal health physics equipment, the necessary meteorological devices must be available for the determination of the point of maximum ground concentration of radioactivity.

7.5.3. A monitoring crew should be trained to carry out a radiation survey during and after an emergency. A good knowledge of the normal radiation levels in the environment should be secured by means of regular surveys.

7.5.4. The monitoring teams should use the best available means of transport and communications for the efficient and speedy collection and transmission of data. The equipment issued for the monitoring crew may be:(a) Radiation measuring instrument with a range between

0.1— 1.0 mr/hr and 1 r/hr; a dust sampler may be useful.(b) Map covering the area with route to be followed and point

where measurements should be taken.(c) Protective clothing (coveralls), overshoes, film badges and

pocket dosimeters, range up to 50 r.(d) Paper, pencil and hand torch.

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A C K N O W L E D G E M E N T

The Director General gratefully acknowledges the work o f the follow ing experts, who formed an international panel on “ Safe Operation of Critical Assemblies and Research Reactors” . The panel held two series o f meetings in Vienna, in February and July 1960.

Chairman:

D. W. Jefterson-Loveday United Kingdom Atom ic Energy Authority United Kingdom

Members:J. Aleksandrowicz Instytut Badan Jadrowych Poland

R. de M ello Cabrita::'Laboratory for Nuclear Physics andEngineeringPortugal

R. Pacheco de Figueiredo Laboratory for Nuclear Physics and Engineering Portugal

S. G. KaufmannArgonne National LaboratoryUnited States of Am erica

N. Lakshmanachar **Atomic Energy Establishment India

N. B. Prasad *Atom ic Energy Establishment IndiaE. O. RoxinComision Nacional de EnergiaAtomicaArgentinaS. SuguriJapan Atom ic Energy ResearchInstituteJapanJ. F. TschernilineInstitute for Atomic EnergyAcadem y o f ScienceUnion of Soviet Socialist RepublicsObservers:Y. van der Feer*Reactor Centrum Nederland Netherlands

R. K. ThomasInternational Organization for Standardization

Scientific Secretary:D. N ewbyInternational Atom ic Energy Agency

The manuscript of the Manual was prepared jointly by D. N ewby (Division of Reactors), Milan Osredkar (Division of Reactors) and G. Jenssen (Division of Health, Safety and Waste Disposal).

Attended first meeting only. Attended second meeting only.


A P P E N D I X I

B I B L I O G R A P H Y

This Bibliography is divided into three sections: General, Reactor Safety Reports, and Siting and Containment. It is a selected list of references, and is in no way to be regarded as a complete list. The omission of any paper or book does not imply in any manner its unsuitability for inclusion.

Unless otherwise stated all references are in English.The following abbreviations have been used:A E C D : United States Atom ic Energy Commission, “ de-classified” reportsA E C L : Atom ic Energy of Canada Ltd., Chalk River, Ont.AECU : United States Atom ic Energy Commission, “ unclassified” reportsAERE: Atom ic Energy Research Establishment, Harwell, EnglandA M F: American Machine and Foundry CompanyA N L : Argonne National LaboratoryBM I: Batelle Memorial InstituteID O : Idaho Operations Office (of USAEC)K: Union Carbide Nuclear Company (Gaseous Diffusion Plant)KAPL: Knolls Atomic Power Laboratory (General Electric Co.)M IT : Massachusetts Institute of TechnologyNAA: North American Aviation, Inc.O RN L: Oak Ridge National LaboratoryTID : Technical Information Service Document, United States Atom ic

Energy Commission

S E C T I O N A General references

U .S. RESEARCH REACTO R OPERATION A N D USE. J. W . Chastain (Addison, W esley Publishing Co. Inc., USA (1958)).(A very com plete manual on this topic, covering all aspects.)

W H A T ’S A V AILABLE IN TH E UNCLASSIFIED A TO M IC ENERGY LITERATU RE. TID-4550, USA (1958).(O f value for locating sources o f reports. Also describes USAEC bi-monthly publications, Nuclear Science Abstracts.)

SE LECTED BIBLIOGRAPHIES ON REACTO R SAFETY. R. J. Smith. T ID - 3073 and TID-3525 (Rev. 1), USA (1958,1959).

N U CLEAR SAFETY (A quarterly Technical Progress Review), ORNL, USA (1959). 1 1, published in Sept. 1959, subsequent issues published quarterly. (A general review o f the field o f nuclear safety.)

N U CLEAR REACTO R EXPERIM ENTS. J. B. Hoag (published bv D. Van Nostrand & Co., USA (1958)).(Describes experiments perform ed at the International Reactor School at Argonne National Laboratory. The sections concerning operation of reactors and the Argonaut reactor are useful.)

SECO N D U N ITED NATIONS IN TE RN ATIO N AL CO N FE REN CE ON TH E PEA CEFU L USES OF A TO M IC ENERGY — Vol. 10, Research Reactors (1958). (Available in English, French and Spanish.)

S2


(Contains many useful references on design, construction and operation of research reactors. Note especially: P/1524, Jackson (UK), Som e considerations of the safety of the D ido C lass Reactors (10 M W D 2O Tank type).)

SECO N D U N IT E D NATION S IN TE R N A TIO N A L CON FE REN CE ON TH E PEA CEFU L USES O F A T O M IC ENERGY — Vol. 11, Reactor Safety and Control (1958). (Available in English, French and Spanish.)(Contains many useful references. Note especially: P/2407, Beck et al. (USA), Reactor Safety H azards Evaluation and Inspection (a description of USA practice); P/436, Gom berg (USA), Risk Evaluation of Reactor Siting (applied in this instance to a large power reactor).)

TE ST REACTO RS M E E T IN G FOR INDUSTRY. IDO-16520, USA, Idaho Falls, Idaho, USA (May 1959).(General and specific papers mainly concerned with large test reactors. Part I — Construction and operation of test reactors. Part II — Utilization of test reactors. Note especially: D . R. de Boisblanc, Reactor and Experiment Safety Considerations, 155-162.)

ATT I D E L CONGRESSO SCIEN TIFICO , Vol. I, Sezione Nucleare, Italy, Giugno 1959.(Proceedings o f Sixth International Congress and Exhibition o f Electronics and Atom ic Energy, Rome, June 1959. Includes papers by : Nicholls (UK), Adm inistrative procedures for ensuring nuclear safety in research establishments, 441-452; Dunster and Farmer (UK), Em ergency Organization and Reactor Siting; Dozinel and Storrer (Belgium), Safety Aspects of the Relgian BR-3 Reactor, 469-529.)

KERN TECHN IK. Riezler-W alcher (published by Teubner, F. R. o f Germany (1958), in German).(Chapter 3.8. is a technical discussion of reactor safety, being part o f a very com plete reference book on reactor physics and technology.)

N U CLEAR SAFETY G U ID E. Callihan et a l , TID-7016, USA (1958).(Contains information pertinent to prevention of accidental criticality during transport, storage and chemical processing o f enriched U235.)

C R IT IC A L EXPERIM EN TS A N D N U CLEAR SAFETY AT OAK RIDGE N ATIO N AL LABORATORY. A. D. Callihan, ORNL-2087, USA (1956).

(A short review o f O RN L experience and practice.)A SUM M ARY OF ACCID EN TS A N D IN CID EN TS IN V O LV IN G R A D IA

TIO N IN A TO M IC ENERGY ACTIVITIES, 1945 to 1955, 1956 and 1957-1958. Daniel F. Hayes, TID-5360, TID-5360 Supplement 1, and TID-5360 Supplement 2, USA (1956, 1957 and 1959).(A good summary o f all incidents which have occurred in USA, including those with reactors where personnel have been exposed to ionizing radiation.)

BOOBY TRAPS. H. Paxton, AECD-4240, USA (1957).(A discussion o f accidents which have occurred in critical assemblies due to unforeseen situations.)

C O N TA M IN A TIO N OF TH E NRU R E A C T O R IN MAY 1958. J. W . Greenwood, AECL-850, Canada (1959).(Describes a fuel element failure in a large D 2O research reactor, and subsequent decontamination operations.)

“ A R G O N A U T” IN TE RIM REPORT A N D EN G IN EER IN G C O N STRU CTIO N A N D COSTS. Lennox and Spinrad and Armstrong et al., ANL-5552 and 5704, USA (1956, 1957),

6* 83


(Description and detailed information on the Argonaut low -pow er research reactor. See also ANL-5647 in section B.)

OPERATIN G M AN U A L FO R THE A R G O N A U T REACTOR. ANL-6036, USA (1959).(Gives details o f personnel organization and operational procedures.)

CO N STRU CTIO N A N D O PERATIO N O F TH E AE 6 W A T E R BO ILER REACTO R. Swanson et al., NAA-SR-1920, USA (1957).(A description o f a 2 kW homogeneous research reactor, together with operating experience, critical approach procedures and operating rules.)

DER FO RSCH U N G S-REACTO R M UNCH EN (The Munich Research Reactor). M aier-Leibnitz (Editor), (in German.) Publisher Karl Thiemig, KG, Munich,F. R. o f Germany (1958).(A general description o f an IM W Swimming Pool Reactor, including buildings, laboratories and utilization.)

CO N STRU CTIO N AN D OPERATION O F TH E RESEARCH REACTO R D R 1. P. Frederiksen, RISO-IO, Denmark (1959). (Danish Atom ic Energy Commission, Riso, Roskilde.)(Description o f Staff Organization, and Operating Experience referred to a 2 k W homogeneous -water boiler reactor.)

OPERA TIN G M AN U AL FO R TH E CP 5 REACTOR. AECU-3526, USA (1956). (The operating manual o f a large 5-M W D fO tank-type research reactor at ANL.)

RESEARCH REACTORS, Parts 1 and 2. (Nuclear Engineering, 5 46 and 47, UK (1960), 99-104 and 154-160.)(A descriptive survey o f research reactors commercially available, mainly from USA and UK.)

ATO M IC PO W E R SYMPOSIUM. AECL-799, Canada (1959), Chalk River, Ontario, May 1959.(N ote paper by G. C. Laurence, Safety in the U se of N uclear Reactors.)

PRE VEN TIO N A N D H A N D LIN G O F R A D IA T IO N EM ERGEN CIES. C. R. Milone, K-1436, USA (1959).(M uch information on administrative requirements.)

TH E O RG AN IZATIO N A T AERE H A R W E L L FOR SAFETY CLE ARAN CE OF REACTORS. Stewart and Wilson, AERE-R-2947, UK (1959).(Describes the organization used at a large nuclear research establishment.)

H ISTORY AN D O PERATIN G PRACTICES OF TH E M TR REACTO R SAFEG U ARD C O M M ITTEE. D . R. de Boisblanc, IDO-16284, USA (1956). (Describes the operations o f a safety committee for a very large test reactor.)

RELIABLE REACTO R PRO TECTIO N . E. Siddall, AECL-430, Canada (1956).(Describes duplication-coincidence principle o f instrumentation.)

R E A C T O R KINETICS — A BIBLIOGRAPHY. N. Bloomfield and F. G. Ben- net, NAA-SR-3808, USA (1959).(Contains some technical references to safety.)

SAFETY PRINCIPLES FO R L O W -P O W E R RESEARCH REACTORS. (N uclear Engineering, 5 46, UK (1960) 96-98), Eltham and Hicks.(Discusses design principles relating mainly to low -pow er research reactors, including that o f reactivity limitation.)

84


SE C TIO N BSafety reports

SUM M ARY REPORT ON TH E H AZARD S OF TH E A R G O N A U T R E A C TOR. Lennox and Kelber, ANL-5647, USA (1956).(Refers to a low power research reactor.) (See also Section A, ANL-5552, 5704 and 6036.)

H AZARD S SUM M ARY REPORT FO R IN D U STR IA L R E A C T O R LAB ORATORIES INC. AM F-G R-10-56 and AM F-GR-10-56 Suppl., USA (1956), b y American Machine and Foundry Co., Greenwich, Conn., USA. (Referring to a pool-type reactor at Plainsboro, N ew Jersey.)

FIN A L H AZARD S SUM M ARY REPOR T T O TH E ADVISORY CO M M ITTE E ON R E A C T O R SAFEGUARDS ON A RESEARCH R E A C T O R FO R TH E M ASSACHUSETTS IN STITU TE O F TEC H N O LOG Y. T. J. Thompson, M. Benedict et al., MIT-5007, USA (1956).(Referring to a 1 M W DgO moderated tank-type reactor located in a highly populated district.)

H AZARDS SUM M ARY REPOR T FO R B A T TE LL E RESEARCH REACTOR. J. W . Chastain, R. F. Redm ond et al., BM I-ACRS-601 (Rev), USA (1955). (Referring to a 1 M W pool-type reactor.)

H A ZA R D SUM M ARY REPORT ON TH E O XID E C R IT IC A L EXPERIM ENTS. W . C. Redman, J. A. Thie and L. R. Dates, ANL-5715, USA (1957). (Referring to a D20-m oderated, heterogeneous, oxide-fuelled system.)

EN G IN EER IN G D ESIGN A N D SAFEGUARDS REPORT O F TH E E N G IN EERIN G TEST REACTO R. Bush et a l , IDO-24020 (TID-4500), USA (1956).(A com plete description and safety report o f a very large (175 M W ) test reactor.)

SICH ERH EITSBERICH T FUR D E N BERLIN ER FORSCHUNGSREAKTOR, BER; und ER G A N ZU N G ZU M SICH ERH EITSB ERICH T (Safety Report for Berlin Research Reactor and Supplement). IKB-C-B 2 and H M I-B 6, F. R. o f Germany (1958 and 1959). Hahn-Meitner-Institut fur Kemforschung, Berlin-W annsee (in German).(A com plete safety report of a 50-kW homogeneous water-boiler research reactor.)

S E C T I O N C Siting and containment

CO N SID ERATIO N S IN TH E CH O ICE OF SITES FO R REACTORS. C .K . Beck, TID-7557, USA (1958).(Paper presented to the Fifth International Congress and Exhibition of Electronics and Atom ic Energy, Rome, Italy, June 1958.)

R EA C TO R C O N TA IN M E N T (Including a technical progress review). R. O.Britten, ANL-5948, USA (1959).

CRITERIA A N D M ETH O D S FO R E V A L U A TIN G E N VIRO N M EN TAL REA C TO R H AZARD S. J.J. Fitzgerald, KAPL-1527, USA (1956).(Criteria and methods for evaluating nuclear hazards are discussed.)

E N V IR O N M E N TA L H AZARD S EVALU ATIO N S FO R C R IT IC A L ASSEMBLIES. J.J. Fitzgerald and R. J. Feinberg, KAPL-1787, USA (1956). (Criteria for site selection are discussed.)

85


ATT I D E L CONGRESSO SC1ENTIFICO, Vol. I, Sezione Nucleare, Giugno1959. Italy (1959).(Proceedings of Sixth International Congress and Exhibition of Electronics and Atom ic Energy, Rom e, June 1959. Includes papers by: Dunster and Farmer (UK), Em ergency Organization and Reactor Siting; (Italy — in Italian), Characteristics of the neighbourhood chosen for the centre of nuclear research at Saluggia, 33-40; D opchie (Belgium), The Incident of the Mol Site on the Safety o f B R 2 , 383-397; Duhamel (France — in French), The choice of sites considering the problem s of radioactive effluent disposal, 253-264; Mas (France — in French), The Id eal Atomic Centre, 105-137.)

See also references under SECTIO N A to papers presented at the Second United Nations International Conference on the Peaceful Uses o f Atomic Energy.

See also the relevent sections o f the reports listed in SECTION B. Most of these reports contain information specific to a particular installation regarding siting and containment.

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A P P E N D I X II

R E A C T O R S T A N D I N G O R D E R S A N D O P E R A T I N G I N S T R U C T I O N S

I t must be noted that this Appendix supplies an example only; it should not be regarded as being applicable to any particular reactor. I t contains typical examples of the application of some of the administrative principles advocated in this Manual to a low-power tank-type reactor. The examples are presented in two sections:SECTION A. Reactor standing orders, issued by the Establishment Direc

tor and laying down the general principles of operation. SECTION B. Operating instructions, issued by the Reactor Manager as

the detailed operating instructions for the plant.

SECTION AReactor standing orders

These orders are issued by the Director of the Establishment for the guidance of organizations and officers concerned with the operation and utilization of the reactor. These orders define the safe operating principles and procedures. The detailed procedures to be followed by operation and maintenance staff will be laid down in the reactor operating instructions by the Reactor Manager.

1 . R e s p o n s ib il it ie s

1.1. The DirectorThe Director is ultimately responsible for the safety of the reactor, in accordance with statutory and other regulations which may be in force.

1.2. Chief ExecutiveThe Chief Executive is responsible for appointing the members of the safety committee, the operations safety committee and the Reactor Manager and will be responsible to the Director for the safe operation, maintenance and utilization of the reactor.

1.3. Safety CommitteeThe Safety Committee will as a body have no executive responsibility, but will act in an advisory capacity to the Director. The Safety Committee will examine the operation and maintenance instructions to advise whether these correctly interpret the principles in the reactor safety documents and whether they fall within the current statutory and other requirements. The Com

87


mittee will inspect the reactor at appropriate intervals to ascertain if it is being operated and maintained in a safe manner and in accordance with the standing orders and operating instructions. The Committee will review the standards and procedures adopted for the design, acceptance and operation of experiments in the reactor. The Committee will receive and consider the reports on all reactor trips and unusual incidents and advise on any appropriate safety action. The Committee will meet on a formal basis and minutes of the Committee’s proceedings will be kept.

1.4. Operations Safety CommitteeThis Committee will have no executive responsibility but will act in an advisory capacity to the management. This Committee will examine all proposed experiments and operations, and advise the management on any aspects which may cause unsafe conditions to arise. The Committee will meet on a formal basis and minutes of the proceedings will be kept.

1.5. Reactor ManagerThe Reactor Manager is responsible to the Chief Executive for the safe operation of the reactor, for the safety and good order of the reactor area and for carrying out the agreed reactor operation programme. He will issue reactor operating instructions for the guidance of his staff and these operating instructions must fall within the framework of the principles laid down in these standing orders. He will issue instructions of a temporary nature in the form of “ temporary operating instructions” which will be valid for a stated period.

1.6. Reactor OperatorsThe reactor operator on duty will be responsible for the safe operation of the reactor within the terms of the reactor operating instructions and within the limits defined by the current operations certificate.

1.7. Maintenance EngineerThe maintenance engineer will be responsible for the efficient maintenance of the reactor plant and equipment. He will issue detailed maintenance instructions and will ensure that no work is begun without the permission of the Reactor Manager or his deputy.

1.8. Reactor PhysicistThe reactor physicist will be responsible to the Reactor Manager

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for maintaining data on the reactivity balance on the reactor, including the effect of core loading and experimental facilities. He must be in attendance at and must take part in all critical approaches on new reactor cores.Reactor ChemistThe reactor chemist will be responsible for the chemical control of the purity of the water in the reactor system.Radiological Protection OfficerThe radiological protection officer is responsible for advising the Reactor Manager on all aspects of radiological protection. He is responsible for:(a) Providing an independent regular check on the radiation and

contamination levels in the reactor area;(b) Providing adequate equipment to enable the operations staff

to carry out a survey on all radiation and contamination hazards;

(c) Operating the health physics aspects of personnel control; and(d) Advising the Reactor Manager on current permissible levels

of radiation exposure and radioactive ingestion and inhalation.

2 . R e a c t i v i t y

The total reactivity of the core must never exceed . . . °/o Ak/k without the written approval of the Reactor Manager.The reactor period trip setting must not be reduced below 10 sec except with the written authority and in the presence of the Reactor Manager.

3 . F u e l l o a d in g

All operations involving the movement of fissile material or absorbers in the reactors core must be carried out under the direct supervision of the Reactor Manager. All alterations to the fuel loading must be completely described and authorized in a coreloading certificate before the operation; the only authorized exception to this is that single elements may be removed and replaced where this is required for the insertion of flux-measuring foils. Full critical approach procedures must be adopted for any change in core configuration; the only authorized exception to this is the replacement of any fuel element in a confirmed core by another element containing the same amount of fissile material. Core changes must only be made if the shut-down rod is fully raised, the control rods are fully inserted, the high-flux trips are


set a t. . . kW, no trip circuits are by-passed and all low-flux instrumentation is working correctly. Changes carried out when a complete core is present must always commence with the removal of fuel with a reactivity worth of at least. .. °/o from the sides of the core.

4 . R e a c t o r o p e r a t i o n

4.1. The reactor must be operated in accordance with the reactor standing orders and the reactor operating instructions.

4.2. The operating intention must be accurately detailed in an operations certificate.

4.3. Operation of the reactor must be strictly in accordance with the current operations certificate, which must be displayed.

4.4. Either the Reactor Manager (or his deputy) or reactor operator must be present at the control console when the reactor is operational or in the stand-by condition, and at least one other person must be in the reactor control room.

5. S h u t - d o w n s t a t e

5.1. Sufficient instrumentation must be operational during shut-down to indicate the state of the reactor and to provide adequate protection. The minimum requirements will be one period channel and one flux-trip channel, plus four health-chamber channels.

6 . S a f e t y c i r c u i t s

6.1. The safety circuits and equipment will not be modified, either temporarily or permanently, without the approval of the Reactor Manager and Safety Committee.

6.2. Only the reactor maintenance staff under the supervision of the reactor maintenance engineer may make modifications to the reactor equipment.

6.3. The Reactor Manager must make arrangements for periodic testing of the reactor safety feature and for acceptance tests for equipment which has been under maintenance.

6 .4 . Interlock Cut-OutsWhen the reactor is operational, interlock cut-outs may only be applied with the approval of the Reactor Manager or his deputy.

7. R e a c t o r p o w e r

7.1. The maximum power at which the reactor is permitted to operateis . . . kW (for a particular core). Calibration of power indication

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channels will be carried out at intervals o f two months or less. Instruments must be recalibrated for a new or different core.

8 .2.

8 .3 .

8 .4 .

8 .5 .

9 .1 .

9 .2 .

10.1.

10.2 .

1 0 .3 .

11 . 1 .

8.1.

11.2.

8. M a in t e n a n c e

Adequate and regular shut-down periods will be arranged for routine maintenance to take place.Maintenance must be carried out in accordance with the schedules prepared by the maintenance engineer.Maintenance on the reactor and associated plant may be undertaken only with the permission o f the Reactor Manager. Maintenance must proceed on only one control or one shut-down rod at one time.All movements of control and shut-down rods must be performed by the properly authorized member of the operations staff.

9 . R a d i a t i o n a n d c o n t a m i n a t i o n c o n t r o l

The Reactor Manager is responsible for the radiation safety of all operations connected with the reactor and for ensuring that the reactor is not operated in a manner hazardous to those in and around the reactor area.All personnel working in the reactor building are required to wear film badges and selected personnel will wear pocket dosimeters. At least one general radiation survey must be made every eight hours during operation.

1 0 . E x p e r im e n t s

All experiments are subject to the approval o f the Reactor Manager. If the safety o f the reactor is involved, or in cases of doubt, advice must be sought from the Safety Committee.The quantities of combustible, inflammable and explosive materials in the reactor building must be kept to an absolute minimum. T he Reactor Manager must be notified of the presence of all such materials.The presence o f noble metals, including silver, copper, mercury and their alloys, in the reactor water system is forbidden.

1 1 . LOSS OF POWER SUPPLIES

The Reactor Manager must be given at least twenty-four hours notice o f the intended withdrawal of pow er supplies.The Reactor Manager or his deputy must be inform ed immediately of any unscheduled loss of pow er supplies.

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1 2 . E m e r g e n c y in s t r u c t io n s

1 2 .1 . Detailed emergency instructions, including instructions for the evacuation o f the building, must b e given in the reactor operating instructions and must be regularly practiced.

1 2 .2 . The reactor operator will report an emergency to the Reactor Manager or his deputy or to other senior officers at the earliest possible moment. A list o f officers to be informed must be available and displayed in the control room.

1 3 . R e a c t o r p r o g r a m m e

1 3 .1 . The reactor operating programme which has been agreed between the reactor users and the operating staff will be displayed in the reactor building.

1 3 .2 . The reactor will generally be available for experimental use at all times, except when under maintenance.

SECTION BReactor operating instructions

These instructions are to be obeyed unless the operation is consideredto be unsafe. In that event the operation should be discontinued, thereactor put in a safe condition and the Reactor Manager notified, in orderthat the fault may be rectified and the programme continued.

1 . R e a c t o r s t a r t -u p p r o c e d u r e

1 .1 . The follow ing documents will be com pleted and in the possession o f the reactor operator before the start-up operation is com m enced:(a) Experiment Loading Certificate; and(b) Operations Certificate.

1 .2 . W ithdraw and cancel all clearance and permit-to-work certificates.1 .3 . Ensure by inspection that the reactor conditions agree with those

stated in the log-book.1 .4 . Ensure that the operations called for on the operation certificate

are fully understood. Check that the reactor conditions are those required for the operations. In particular confirm that the core configuration is as shown on the operation certificate and the control room display board.

1 .5 . Ensure that operating personnel are wearing film badges and that no unauthorized persons are in the reactor building.

9 2


1.6 .

1 .7 .

1.8.

1 .9 .

1.10.

1.11 .

1 . 12.

1 .1 3 .

1 .1 4 .

Switch on instrument pow er supply. Check that “ Reactor Shut- D ow n ” is indicated.Check high-voltage supply for the four counter channels and adjust as required.Set discriminator and gain controls o f the sub-critical and startup channel amplifiers to normal operating positions. The correct settings will be obtained from the Reactor Manager and a record m ade in the log-book.Set start-up channel trip levels. Set linear pow er channel to most sensitive range.Set log pow er and period meter trip level. Set shut-down channel trip level.Obtain pow er key and switch on control power. Check that “ Control Power” is indicated. N ote operation in log-book.Insert reactor master key and turn master selector to “ Source” position. Press source “ In” switch. Check that “ O ut” light is extinguished and source “ In ” indicator is lit within about one minute. (M ovem ent o f source may cause a period trip but this may be avoided by careful use of the “ In” switch. Should a period trip occur, reset the trip and re-energize the control power.) Check that normal readings are obtained on linear instrument and start-up channel. Note operation in log-book.Clear low-level trips.Turn master selector to shut-down rod and press shut-down rod “ raise” switch. Little or no change should occur in instrument reading as a result o f this operation. After about four minutes shut-down rod “ out” should be indicated.Turn master selector to Dum p Valve. Operate D um p Valve switch and check that “ Dum p Valve C losed” is indicated.Turn master selector to “ W ater” . Operate pump switch to admit water to core. Check that “ W ater Pump O N ” is indicated. Note period o f pum ping to reach normal level and watch instruments carefully during this period. Instrument readings should generally indicate a fairly rapid drop initially and then rise to something less than the initial value.Stop pump when “ Normal water level” is indicated.Note operation in log book with details o f time to admit water, water temperature and instrument readings.Turn master selector to “ Central Rods” ; press “ Raise Control R od” switch to approach criticality. W atch instruments carefully.

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Stop rod withdrawal at intervals to observe the indicated period apparent during rod withdrawal. As criticality is approached, the length o f time that this indicated period remains after rod withdrawal is stopped progressively increases until at criticality a small constant period will b e observed. Check that criticality is being achieved with control rods near to the predicted position. If excessive difference is apparent, stop operation and investigate reason.

1.16. Low er control rods to remove the indicated period.1.17. Turn master selector to “ Source” position. Press source “ O ut”

switch. Check that source “ In ” light is extinguished and source “ O ut” indicator is lit within about one minute.

L.18. Raise control rods slightly if instrument readings continue to fallafter source is withdrawn.

1.19. Raise control rods to produce a reactor period of about 20 sec (never less than 10 sec). W hen desired pow er level is approached, lower control rods to critical position, making minor adjustments to hold power constant.Note operation in log-book with details o f source and fine-control rod position, water temperature and instrument readings.

1.20. Continue operation in accordance with programme laid dow n in the operations certificate. Log-book entries will be made at hourly intervals and also when a change is made in operating conditions or significant events occur.

Only reactor start-up procedures have been given here. These must befollow ed by instructions for continued operation, normal shut-down, emergency shut-down, maintenance, emergency procedures, etc., etc.

94


APPENDIX IIISYLLABUS OF A REACTOR T R A I NI NG COURSE

(Refer to Section 4.3.3. for other information regarding this course)

A. Elementary mathematics (10 hours)1. Ordinary differential equations.2 . Differential equations leading to special functions: Bessel, Legen

dre, etc.

B. Atomics and elementary nuclear physics (15 hours)1. Avogadro’s number — sizes and masses of atoms.2 . Properties o f cathode rays.3. Atom ic structure — Atom ic spectra in terms of Bohr’s theory —

spectrum o f hydrogen — shell structure of the atom — periodic table.

4. X-rays — production and detection — properties.5. Radioactivity — properties o f radiations — radioactive series and

their disintegration products — detection — scattering of alpha particles and the sizes o f the nuclei.

6. Isotopes — binding energy.7. Ideas o f rest and relativistic mass and energy — conservation of

mass and energy.8. Nuclear reactions induced by neutrons and other nuclear par

ticles — fission and nuclear energy — transuranic elements.

C. Reactor physics (15 hours)1. Neutron temperatures, neutron flux, cross-sections and nuclear

chain reaction.2 . Diffusion o f neutrons.3. Slowing-down o f neutrons.4. Experimental determination of slowing-dovvn and diffusion

constants.5. Bare homogeneous thermal reactor.6. Introduction to heterogeneous reactors.7. Control o f nuclear reactors.8. Experiments with critical and sub-critioal assemblies — determina

tion o f critical size — use o f a reactor in the study o f Bssion physics, solid state physics and shielding experiments.

D. Reactor engineering (15 hours)1. Engineering features of reactor systems.

95


2. Choice o f coolant, moderator and fuel element.3. Description of selected examples of research and pow er reactors.4. Reactor materials.5. Fuel-element fabrication.6. Application o f nuclear reactors to power generation and its

economics.

E. Electronics and radiation detection (20 hours)1. Fundamentals o f vacuum tubes — static dynamic characteristics.2. Power supply circuits.3. Simple theory of R— C coupled amplifiers and feed-back amplifiers

including the cathode follower.4. Cathode-ray oscilloscope and its applications.5. Multivibrators, pulse generators and scalers.6. Vacuum-tube voltmeters and their applications.7- General theory of energy loss of electromagnetic rays and charged

particles in matter — track density, energy per ion pair, range and mobility of ions and electrons.

8. Gas-filled counters, ion chambers, d. c. ion chambers, proportional counters, BF3 counters, fission counters and Geiger counters, etc.

9. Scintillation counters and photomultipliers.1 0 . Activation method of detection of neutrons.1 1 . Counting statistics and theory of errors.

F. Radiological protection and radiobiology (10 hours)P a r t I — H e a l t h ph y sics

1. Maximum permissible levels of radiation.2. Radiation detection instruments and measurements.3. Shielding calculation for X-rays, gamma-rays, beta-rays and neu

trons.4. Disposal of radioactive wastes.5. Radiation and monitoring of environment, area and buildings.6. External and internal radiation hazards — radiation hazards from

different types o f reactor and control measures.

P a rt II — R a d io b io l o g y

7. A broad survey of the effects of ionizing radiations on living organisms — radiation injury in the context of damage to cells, tissues, organ systems and finally to the whole body.

8. Discussion of the theories concerning the mechanisms of action of ionizing radiations on living systems.

9 6


C. Radiochemistry and isotope handling (10 hours)1. Chemistry o f uranium, thorium, plutonium and other elements of

interest in atomic energy.2. Radioactivity and radioactive decay, nuclear reactions, fission and

activation analysis.3. Radiochemical separation techniques and chemistry of fuel pro

cessing.4. Radiation chemistry and effects of'radiation on solids.5. Preparation and uses o f isotopes.6. Special analytical techniques.

97


O T H E R A G E N C Y P U B L I C A T I O N S I N T H E

S A F E T Y S E R I E S

No. 1 — SAFE HANDLING OF RADIOISOTOPES — STI/PUB/1 100 p. (14 .8X 21 cm) — (B) — US $ 1 ; 6s.0d.stg; Sch 21

This Manual, prepared by an international group of experts in consultation with other international agencies, covers medical, technical and organizational aspects of safety practices and deals with maximum permissible levels for exposure to radiation, organization of safety, medical supervision of workers, monitoring and records, use, storage and transportation of sealed and unsealed sources, accidents, decontamination and waste disposal.Available in English, French, Russian and Spanish.

No. 2 — SAFE HANDLING OF RADIOISOTOPES — HEALTH PHYSICS ADDENDUM — STI/PUB/10120 p. (14.8X 21 cm) — (B) — US $ 1.50; 9s. Od. stg; • Sch 31.50

The Health Physics Addendum is one of two supplements to the Manual on “ Safe Handling of Radioisotopes” , which the IAEA published in 1958. It contains technical information needed by health physicists in implementing the controls recommended in the Manual and was compiled by the Agency’s Secretariat on the basis of material prepared by two of the expert members of the Panel whose recommendations form the text of the Manual itself. A valuable feature of the Addendum is the Annex of useful health physics data in the form of tables, diagrams and illustrations of instruments.Available in English, French, Russian and Spanish.

No. 3 — SAFE HANDLING OF RADIOISOTOPES — MEDICAL ADDENDUM — STI/PUB/1180 p. (14.8X 21 cm) — (B) — U S$1.50; 9s.0d.stg; Sch 31.50

The Medical Addendum is one of two supplements to the Manual on “ SafeHandling of Radioisotopes” , which the IAEA published in 1958. It containstechnical information needed by medical officers in implementing the controls recommended in the Manual and was compiled by the Agency’s Secretariat on the basis of material prepared by two of the .expert members of the Panel whose recommendations form the text of the Manual itself. It also contains a useful bibliography of the relevant international literature.Available in English, French, Russian and Spanish.


No. 5 — RADIOACTIVE WASTE DISPOSAL INTO THE SEA — STI/PUB/14

This document, which has become known as die “ Brynielsson Report” , is the report of a panel convened by the Agency’s Director General and presided over by Mr. Harry Brynielsson, managing director of AB Atomenergi (Sweden). While it does not necessarily express the views on this subject either of the Agency or of the bodies to which the individual panel members belong, it represents the general considered opinion of a group of distinguished scientists and other experts, who believe that waste disposal operations can be controlled in such a way as to safeguard man against the deleterious effects of radiation. A series of recommendations for an international agreement to that effect is offered, together with material for the practical guidance of those who are technically concerned with radioactive waste disposal into the sea.Available in English, French, Russian and Spanish.

No. 6 — REGULATIONS FOR THE SAFE TRANSPORT OF RADIOACTIVE MATERIALS — STI/PUB/40

This publication contains the Regulations for the Safe Transport of Radioactive Materials to be applied to all Agency operations and to Agency-assisted operations. These regulations were prepared in draft form by two panels of experts. Topics dealt with include packaging requirements, limitation of external dose rate, general requirements for the transport of radioactive materials of low specific activity, of fissile materials, and of large radioactive sources. The Annexes to the regulations provide information on the classification of radionuclides in terms of their radiotoxicity and on the methods of ensuring that packages of fissile material are safe from nuclear interaction.Available in English, French, Russian and Spanish.

No. 7 — REGULATIONS FOR THE TRANSPORT OF RADIOACTIVE MATERIALS: NOTES ON CERTAIN ASPECTS OF THE REGULATIONS — STI/PUB/32 (in press)

This booklet contains background information on the Regulations for the Safe Transport of Radioactive Materials (Safety Series No. 6). On the one hand, some of the scientific considerations which led to the specific limits laid down in the Regulations are discussed; on the other, the practical information it contains on the meaning and use of the Regulations, including a “ layman’s guide” , a synoptic table, guidance on packaging, etc., will be of value to transporters of radioactive materials who are called upon to comply with the Regulations.


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